ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 76

DOE Review Release 07/20/2011 

Document Title:

Feed Materials Production Center Internal Dose 

Topics

Document Number: ORAUT-RPRT-0052

Revision: 00

Effective Date: 07/12/2011

Type of Document: Report

Supersedes: None

Subject Expert(s): Karin Jessen, Melton H. Chew, Robert L. Morris, Eugene W. Potter, Samuel L. T. Chu, 

Leo G. Faust, Bryce L. Rich, Paul Ruhter, and Mark Rolfes (NIOSH)

Approval: Signature on File Approval Date: 06/23/2011

Matthew C. McFee, Document Owner

Concurrence: Signature on File Concurrence Date: 07/05/2011

James P. Griffin, Deputy Project Director

Concurrence: Signature on File Concurrence Date: 06/22/2011

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 07/12/2011

Chris Ellison, Acting Deputy Director, DCAS

New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION. 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 2 of 76

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

07/12/2011 00 New report to consolidate the information pertinent to the Feed 

Materials Production Center (FMPC) internal dosimetry that has been 

developed since the SEC Evaluation Report was presented. This 

information includes topic discussed in the Advisory Board FMPC 

Work Group meetings. Several distinct topics are included in this 

report. Incorporates formal NIOSH review comments. Training 

required: As determined by the Objective Manager. Initiated by 

Karin Jessen.
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 3 of 76

TABLE OF CONTENTS

SECTION TITLE

Acronyms and Abbreviations ...............................................................................................................

PAGE

6

1.0 Introduction .............................................................................................................................. 8

2.0 Radon Releases from K-65 Silos ............................................................................................. 8

2.1 Background ..................................................................................................................8

2.2 The Alternative Assessment of Radon Releases from K-65 Silos ................................. 9

2.3 Radon Release Calculation Using Conventional Modeling Approach ......................... 11

2.3.1 Calculation for Radon Release Rate from K-65 Material to the Silo Air ........... 11

2.3.1.1 Radon Exhalation Rate from K-65 Material in the Silos..................... 11

2.3.2 Calculation for the Radon Rate of Release to the Atmosphere........................ 13

2.3.2.1 CONTAIN 2.0 Leak Path Factor Calculations ................................... 15

2.3.2.2 CONTAIN Code Results ................................................................... 17

2.3.3 Environmental Radon Release Rate ............................................................... 17

2.3.4 Summary Conclusion...................................................................................... 19

2.4 Response to April 2010 SC&A White Paper ...............................................................19

2.5 Conclusion..................................................................................................................22

3.0 Q-11, K-65, Domestic Uranium, and Raffinates...................................................................... 23

3.1 Processing History of Pitchblende and Domestic Uranium Feedstock ........................ 23

3.2 Available Monitoring Data...........................................................................................25

3.3 Recommendations......................................................................................................31

3.4 Detailed Timeline and History of K-65 Material ...........................................................31

4.0 Recycled Uranium at FMPC................................................................................................... 33

4.1 Historical Background and Process Limits..................................................................33

4.1.1 Major Contaminants in FMPC RU ................................................................... 34

4.1.2 Mass Balance, Material Flow, and Pu-239 Levels in RU within the 

Weapons Complex.......................................................................................... 34

4.1.3 Other Contaminants in RU .............................................................................. 36

4.1.4 Tower Ash and Plutonium Out of Specification................................................ 37

4.2 Workplace and Personnel Monitoring Associated with RU.......................................... 38

4.2.1 In Vivo Analysis Data ...................................................................................... 39

4.2.2 Urine Analysis Data ........................................................................................ 39

4.2.3 Air Sampling Data........................................................................................... 40

4.3 RU in Waste Streams ................................................................................................. 41

4.3.1 Process Raffinates and Other Wastes Below the Economic Disposal 

Limit (EDL)...................................................................................................... 41

4.3.2 Daily Weighted Exposure (DWE) Studies........................................................ 41

4.3.3 Other Factors Associated with Recycled Uranium in Waste Streams.............. 42

4.4 RU Mass Flows and Discrepancies in Mass Flow Reports ......................................... 42

4.5 Summary and Recommended Default Values for RU at FMPC .................................. 44

5.0 Use of Daily Weighted Exposure Reports for Estimation of Chronic Daily Thorium 

Intake Rate ............................................................................................................................ 45

5.1 Daily Weighted Exposure Reports ..............................................................................45

5.2 Correlation of Data and Urinary Uranium Excretion Rate............................................ 48

5.3 Uncertainty .................................................................................................................51
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 4 of 76

5.4 Selecting a Daily Weighted Exposure or Air Sample Concentration Value for 

Intake Rate Calculation...............................................................................................51

5.5 Calculation of Chronic Daily Intake Rates...................................................................52

5.5.1 Calculation of Inhalation Rate ......................................................................... 52

5.5.2 Calculation of Daily Ingestion Intake Rate....................................................... 53

6.0 Attributions and Annotations .................................................................................................. 54

References ........................................................................................................................................ 55

ATTACHMENT A, CONTAIN CODE CALCULATIONS AND RESULTS............................................ 62

ATTACHMENT B, GAS DATA APPLIED IN CONTAIN MODEL ........................................................ 69

ATTACHMENT C, WIND MODELING IN THE CONTAIN MODEL .................................................... 70

ATTACHMENT D, VENT GEOMETRY CALCULATIONS.................................................................. 71

ATTACHMENT E, PROBLEM INPUT DECK..................................................................................... 72

LIST OF TABLES

TABLE TITLE

1 Isotopic composition in FMPC Silos 1, 2, and 3......................................................................

PAGE

24

2 DWE data for Plant 2, 1955 to 1962, when Silos 2 and 3 were being filled............................. 26

3 DWE data for Plant 3, 1955 to 1962, when Silos 2 and 3 were being filled............................. 27

4 DWE data for Plant 2 and 3, 1965 to 1968, during low acid processing period....................... 28

5 Radon breath analysis statistical parameters ......................................................................... 29

6 Ratios of Silo 1 and Silo 2 thorium isotopes to 226Ra .............................................................. 30

7 Factors for unmonitored isotopes for the period 1953 through 1962 only ............................... 31

8 Materials placed in Silo 2 after Q-11/K-65 processing was complete ..................................... 32

9 Inventory of K-65 material in Silos 1 and 2 ............................................................................. 32

10 Average characteristics of RU processed at FMPC................................................................ 34

11 Activity in RU containing 10 ppb and 100 ppb 239Pu for various uranium 

enrichments ...........................................................................................................................34

12 Limits for other contaminants in RU ....................................................................................... 36

13 Other radioactive contaminants in RU at Hanford .................................................................. 37

14 RU contaminant levels by process subgroups........................................................................ 37

15 Derived air concentrations from 1989 RU isotopic analysis .................................................... 40

16 Processes that concentrated RU contaminants...................................................................... 41

17 Examples of uranium mass balance inconsistencies in FMPC receipts.................................. 43

18a Recommended RU default values beginning in 1961and to 1973 .......................................... 44

18b Recommended RU default values beginning in 1973 and to 1986 ......................................... 44

19 Where and when thorium was processed at FMPC before 1968............................................ 45

20 Comparison of uranium-in-urine concentration with nominal DWE......................................... 50

21 DWE values, in multiples of the MAC, recommended for use in estimation of 

thorium intake rates................................................................................................................52

A-1 Vent flow velocities versus time.............................................................................................. 63

A-2 Flow from the engineered vents versus time for Problem M2-D. ............................................ 64

A-3 Mass of CO2 vented to the environment versus time for Problem M2-D ................................. 65

A-4 Atmospheric molar fractions................................................................................................... 67

A-5 CO2 vented versus molar fraction CO2.................................................................................... 67
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 5 of 76

LIST OF FIGURES

FIGURE TITLE

1 K-65 silos at the FMPC ..........................................................................................................

PAGE

15

2 Flows to the environment from the engineered vents ............................................................. 17

3 Flows to the environment from the engineered vents ............................................................. 18

4 Percent of the CO2 vented to the environment versus time .................................................... 18

5 RU mass flow in and out of FMPC ......................................................................................... 35

6 Excerpt from a 1955 Plant 9 report showing DWE for 119 employees by job 

description .............................................................................................................................47

7 Excerpt from the appendix of a 1955 Plant 9 report showing operations 

associated with a job description and the data used to assess each operation ...................... 48

8 Uranium-in-urine excretion rate as a function of DWE............................................................ 50

A-1 Vent flow velocities versus time.............................................................................................. 62

A-2 Flows to the environment from the engineered vents ............................................................. 64

A-3 Mass of CO2 vented to the environment versus time.............................................................. 65

A-4 Percent of the CO2 vented to the environment as a function of time....................................... 66
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 6 of 76

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AMC African Metals Company

AWE Atomic Weapons Employer

Bq becquerel 

cfm cubic feet per minute

CFR Code of Federal Regulations

Ci curie

d day

DAC derived air concentration

DCAS Department of Compensation Analysis and Support

DOE U.S. Department of Energy

dpm disintegrations per minute

dps disintegrations per second

DWE daily weighted exposure

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

FMPC Feed Materials Production Center

ft feet

g gram

gal gallon

GDP gaseous diffusion plant

HASL Health and Safety Laboratory

hr hour

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

in. inch

K Kelvin

kg kilogram

L liter

lb pound

LEU low-enriched uranium

LOOW Lake Ontario Ordnance Works

LPF leak path factor

m meter

MAC maximum allowable concentration

MCW Mallinckrodt Chemical Works

MDA minimum detectable amount

MIVRML Mobile In Vivo Radioactivity Measurements Laboratory

mL milliliter

MPC maximum permissible concentrations

mph miles per hour
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 7 of 76

MPLB maximum permissible lung burden

MT metric ton

MTU metric ton of uranium

nCi nanoCurie

NIOSH National Institute for Occupational Safety and Health

NLO National Lead of Ohio

NU natural uranium

ORAU Oak Ridge Associated Universities

pCi picocurie

PGDP Paducah Gaseous Diffusion Plant

POOS plutonium out of specification

PORTS Portsmouth Gaseous Diffusion Plant

ppb parts per billion

ppm parts per million

PUREX plutonium-uranium extraction

RAC Radiological Assessments Corporation

RFP Rocky Flats Plant

RGM radon gas monitor

RMI Reactive Metals, Inc.

RU recycled uranium

s second

SC&A S. Cohen & Associates

SEC Special Exposure Cohort

SRDB Ref ID Site Research Database Reference Identification (number)

SRS Savannah River Site

t ton (2,000 lb)

TASL Track Analysis Systems, Ltd.

TRU transuranic

TWA time-weighted average

UNH Uranyl nitrate hexahydrate

W watt

wk week

yr year

°F degrees Fahrenheit

μCi microCurie

μg microgram
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 8 of 76

1.0

The purpose of this report is to consolidate the information on internal dosimetry at the Feed Materials 

Production Center (FMPC) that has been developed since the Special Exposure Cohort (SEC) 

Evaluation Report was presented (NIOSH 2006). This information includes topics discussed in the 

Advisory Board FMPC Work Group meetings. Several topics are included in this report; each stands 

alone in its own section. The following topics are included in this report: 

INTRODUCTION

• Radon releases from K-65 silos (Section 2.0), 

• Exposure to Q-11, K-65, domestic uranium, and raffinates (Section 3.0), 

• Recycled uranium (RU) (Section 4.0), and

• Use of daily weighted exposure (DWE) data to determine thorium intake rates (Section 5.0). 

In addition, the National Institute for Occupational Safety and Health (NIOSH) has proposed a 

coworker model to describe the intake rates of thorium based on chest count data. That model, which 

is applicable from 1968 through 1988, is not included in this report but is expected to be published at 

a future date.

Attributions and annotations, indicated by bracketed callouts and used to identify the source, 

justification, or clarification of the associated information, are presented in Section 6.0. 

2.0

2.1 BACKGROUND 

RADON RELEASES FROM K-65 SILOS

Original approach defining the amount of radon released from the K-65 silos is defined in the site 

profile (ORAUT 2006b). In that document 

In November 2008, Sanford Cohen & Associates (SC&A) prepared a white paper, An Alternative 

Assessment of Radon Releases from K-65 Silos (Alternative Assessment; SC&A 2008) that provided 

assumptions and quantitative calculations to demonstrate its claim of the potential for 222Rn releases 

between 64,500 to 92,000 Ci/yr from Silo 1 at FMPC. In response to the Alternative Assessment, 

NIOSH prepared Response to SC&A White Paper, An Alternative Assessment of Radon Releases 

from K-65 Silos in November 2008 (NIOSH 2008), which stated:

NIOSH estimated a 222Rn release rate of 5,000 to 6,000 

Ci/year. 

NIOSH disagrees with the draft findings reported by SC&A. The RAC [Radiological 

Assessments Corporation] model was supported by a National Academy of Sciences 

review; SC&A’s was not. NIOSH will rely upon the individual exposure estimates 

produced using the Pinney/Hornung model, which utilized radon exposure levels from 

the RAC study, plus an additional radon source term at Fernald.

Dr. John Mauro made the following statement in the Work Group Meeting on January 29, 2010 

(ABRWH 2010, pp. 280-281): 

Please do one other thing for us. We certainly will look at those papers very carefully, 

and let's say we walk away from this. You read those papers and it looks pretty good, 

and then you take a look at our papers, at the arguments based on ... this deficit of 

lead-210, and you say, then where did the radon go? I mean, we've got ourselves 

quite a dilemma here because the radon had to go somewhere because of that deficit. 

Now, if we could somehow reconcile whatever you have here and our analysis, ... it 

would make me very happy that we could somehow reconcile how one analysis, which 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 9 of 76

is based on the radon deficit (our work), is coming up with some number, and I will look 

at the other work that was done and how they came up with theirs.

Dr. Mauro reiterated this desire in his Division of Compensation Analysis and Support (DCAS)/SC&A 

action items status summary listing (Mauro 2010): 

Issue 5: Review of radon emissions from the K-65 silos

...during the meeting, the work group requested that NIOSH explicitly address the Pb210 deficit arguments, and the external dose measurement arguments made by SC&A 

regarding this matter.... 

In April 2010, SC&A issued A Second White Paper Addressing Enhanced Radon Releases from the 

K-65 Silos at the Fernald Site (Second White Paper; SC&A 2010) in response to NIOSH’s rejection of 

the Alternative Assessment. 

In May 2010, NIOSH prepared a response (NIOSH 2010) to the Second White Paper in which it 

responded to the stated issues on the same subject.

The manner by which K-65 radon release issues were raised as indicated above has resulted in the 

need to consolidate information in one document to provide a more cohesive and comprehensive 

response. To this end, this section focuses on responding to (1) the calculation methodology by 

which the enhanced radon releases were derived in the Alternative Assessment (SC&A 2008), 

(2) calculation modeling based on physical phenomena, and (3) the salient points made in the second 

White Paper (SC&A 2010) in relation to NIOSH’s dose reconstruction position to “rely upon the 

individual exposure estimates produced using the Pinney/Hornung model, which utilized radon 

exposure levels from the RAC study, plus an additional radon source term at Fernald” (NIOSH 2008).

2.2 THE ALTERNATIVE ASSESSMENT OF RADON RELEASES FROM K-65 SILOS

The Alternative Assessment (SC&A 2008) is based on the premise that the degree of disequilibrium 

between 226Ra and 210Pb in the K-65 waste medium determines the amount of radon released in the 

silo headspace. The 210Pb and 226Ra contents in the waste package were based on results from the 

radionuclide analyses of samples taken from the silos in 1991. The amount of radon released to the 

atmosphere before 1979 was based on the relationship of external dose rate on the top of the dome 

taken a minimum of 14 years apart. 

The cumulative uncertainties of the bases for the quantities of radon release in the Alternative 

Assessment (SC&A 2008) are sufficiently large that they invalidate use of the derived radon releases. 

The following list identifies some of the prominent uncertainties that challenge the derivation of the 

release quantities in the Alternative Assessment: 

1. The range of 226Ra concentrations in the 1991 sampling of the silos had shown that the K-65 

material is not homogeneous. “The color of the material also varied greatly,” stated RAC 

(1995). The samples were taken through the four former influent manholes on top of each 

silo, which are denoted by their direction from the center of the Silo (NE, SE, SW, and NW). 

At best, samples from each manhole would cover one-fourth of the K-65 materials. Table 2, 

“Results of Radionuclide Analyses on K-65 Material Obtained by ASI/IT from 1991 Sampling”

in SC&A (2008) presented results of the samples from three of the four manholes (SE, NE, 

and NW). Implicitly, the radionuclide contents of a quarter of the K-65 materials were not 

measured. Confirmed by the fact that the K-65 material is not homogeneous, this introduces 

significant uncertainty to the radionuclide dataset for all K-65 materials.
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 10 of 76

2. The determination of radon release based on the argument of 210Pb deficit did not take into 

account the removal of lead during the chemical processing of the ore and handling of the 

raffinates. It is certain that the 210Pb could have plated out on the surface of the K-65 

materials, as well as on the foam layer on top of the K-65 materials and on the inner concrete 

dome surfaces. On further examination of Table 2 in SC&A (2008), the mean 210Po content 

was found to be 242,000 pCi/g. This amount exceeded the mean 210Pb activity of 

194,000 pCi/g, and must mean that 210Pb is still in the silos somewhere, because there is an 

excess of 210Po. 

3. There is no evidence that the 1991 measured radionuclide contents can be extrapolated to the 

early years of the silos. As a point of reference, Table J-4 of RAC (1995) shows that 

significant variations existed in measurements from different periods. The activity 

concentration of radionuclides that have different solubility would be likely to change because 

of evaporation over time. It is predictable that greater differences would occur in the early 

years when the silo contents were in the process of settling down from previous loading 

operations. 

4. The approach to relate measured external dose rates at the top of the dome to demonstrate 

buildup differences in the headspace is so uncertain as to be unusable. The Alternative 

Assessment (SC&A 2008) failed to ensure that the same equipment was used to measure the 

external dose rates on the silo domes. It is unlikely that the same dose rate meters were used 

from 1964 through 1987. The paper did not substantiate that the measurements were 

conducted in a highly proceduralized manner such that they were taken in the exact same 

geometry and in the exact same location on the silo dome. At least eight or nine different 

individuals took the dose rate measurements over the 23-year period, and it is highly unlikely 

that the measurements were taken in identical locations on the silo dome with the same 

geometries. In addition, the geometry of the silo dome varied over time. In fact, the concrete 

domes on the silos are thinnest in the center and thickest around the perimeter. By the 1970s, 

the degradation of the dome had advanced to a point where the inside of the concrete dome 

was spalling and likely resulted in a much thinner dome. The thinning of the concrete dome 

would dramatically decrease the gamma shielding and account for the observed increase in 

dose rates.

5. Section 3.4 of the report provides the basis for an estimate of 226Ra activity in K-65 Silos 1 and 

2. The exact activity of 226Ra in the two silos is not known for certain, but the amount of 

activity is in the range from 1,650 to 4,600 Ci. The 226Ra activity in Silo 3 is expected to be 

less since the concentration in the Silo 3 material was lower, bringing the total to nominally

5000 Ci or less. 226Ra has a relatively long half-life of 1600 years so the radium activity is 

essentially unchanging during the period of concern. The activity of 222Rn is controlled by, and 

cannot exceed, the activity of 226Ra. Therefore the activity of 222Rn that is available to release 

is nominally 5000 Ci or less, which is in the range of values accepted by NIOSH, and much 

less than the range of values suggested in SC&A’s alternate assessment. 

Section 2.3 presents a simplified radon release rate calculation using a conventional modeling 

approach to illustrate the conservative nature of the RAC (1995) analytical method, for which the 

National Academy of Sciences review did not reveal technical issues that could lead to the prohibition 

on the use of its results.
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 11 of 76

2.3 RADON RELEASE CALCULATION USING CONVENTIONAL MODELING APPROACH

2.3.1

To reach the headspace of the silos, the radon from disintegration of 226Ra nuclides in the waste 

medium must (1) escape from the 226Ra particle of original residence to the pores of the medium 

through release of the kinetic energy of formation (recoil energy) and (2) migrate up through the 

medium materials. At the same time, the radioactive decay process removes quantities of radon 

depending on the time for the radon to migrate through about 20 ft of soil-like materials to the 

headspace. 

Calculation for Radon Release Rate from K-65 Material to the Silo Air

The following is a description of radon transport through soil: Emanation is a several step process – 

first, the radon must escape from the particle in which the parent radium is contained through 

expending the kinetic energy of formation (recoil energy). Alpha recoil is defined as the process by 

which an atom (radon) recoils in the opposite direction from the path of particle ejection following the 

radioactive decay of its parent atom. Most of the radon atoms remain in the particle matrix without 

ever entering the pores in the medium. 

Second, the process continues as radon is transported from a solid to a gas or liquid medium within 

the interstitial pores of the medium, i.e. after radon is produced at the soil particulate level from the 

radioactive decay of radium, it is released into small air or water containing pores between soil and 

rock particles. This transportation of radon through soil is primarily accomplished by the mechanical 

flow of air and/or water through the soil.

Third, after radon is released into the pore spaces, the efficiency of its eventual release into ambient 

air [in the case of the K-65 silos, the silo headspace], termed exhalation, is a function of:

The soil porosity

The concentration of radon in the soil/gas pore

a. Meteorological factors, including, precipitation and atmospheric pressure.

The K-65 silo radon production rate is generally constrained by existing radon in the air spaces. 

However, to ensure favorability to the claimant, an unconstrained production rate (no radon to inhibit 

release from the K-65 material) is calculated in the next section. 

2.3.1.1 Radon Exhalation Rate from K-65 Material in the Silos

The exhalation rate for radon permeated from the soil can be estimated based on the conservation of 

mass; an idealized, one-dimensional, steady-state model for the transportation and distribution of 

radon in the soil can be expressed as the following differential equation (Jasailtis and Girgzdys 2007): 

De

d2C

dx2 − λλλλ +

A

pef

= 0

(1)

where

De = the effective radon diffusion coefficient (m2/s)

C = the radon concentration in pore air of the soil (Bq/m3)

x = a distance from the ground surface with its positive direction downward (m)

λ = the radon decay constant (1/s)

pef = the effective porosity of the soil

A = the production rate of radon gas in the pore space; defined as Equation 2:
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 12 of 76

AA = λλ pp RR EE (2)

where

ρ = the dry bulk density of the soil (kg/m3)

R = 226Ra activity concentration in the soil particles (Bq/kg)

E = the radon emanation coefficient

Given the boundary conditions when x = 0, C = C0, and x→∞, C is finite, the solution for Equation 1. 

(3)

On the other hand, as radon flux is continuous at the ground surface between the soil and 

atmosphere, the exhalation rate F of radon from the ground surface can be expressed as follows:

(4)

Combined with Equation 3, Equation 4 can be rewritten as: 

(5)

The effective radon diffusion coefficient De in the soil has been experimentally studied, which can be 

expressed as the following equation:

DDee = ppDDOO exp(−6mm − 6mm14pp ) (6)

where

p = the total soil porosity 

DO = the radon diffusion coefficient in open air with a constant of 1.1 × 10–5 m2/s

m = the volume fraction of water saturation which can be calculated from water mass 

content ω by using Equation 7:

mm = ωωωω⁄1000 (7) 

The above equations indicate that the radon flux rate can be calculated when the parameters of R, ω, 

ρ, p, and E in the soil are known.

Parameter Values Used in the Calculations

The relevant characteristics of the K-65 raffinate can be found in Appendix J of RAC (1995) and are 

repeated in the following along with the values selected to be used for this calculation.

No specific values were reported for the porosity of the K-65 material. However, the 

report of the 1989 sampling (DOE 1990) reports specific gravity for eight samples (of 

which two are composites) to be between 2.58 and 3.37, with mean 2.98 and standard 

deviation 0.29 (about 10%). Porosity can be calculated from bulk density and specific 

gravity............results in relatively high (compared to typical uranium mill tailings) 

nominal estimates of porosity from about 0.6 to 0.8 (RAC 1995, p. J-9).
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 13 of 76

Porosity p of 0.8 is used for the K-65 material.

The range of moisture contents reported for the 1989 sampling is a summary of eight 

measurements (DOE 1990). Of the eight, five were 20-35%, one was roughly 50%, 

and two were between 70% and 75%. It seems likely that the two highest values were 

for saturated material (RAC 1995, p. J-9).

Moisture content ω of 15% is used for the K-65 material.

Table J-4 have been converted to dry densities, using the moisture content (65% dry 

weight) calculated from information on the related laboratory analytical data sheet 

(NLCO 1972). The calculated dry bulk densities of about 0.53 to 0.72 g cm-3 and the 

value of 1.179 g cm-3 (basis unknown) seem quite low, relative to a more typical value 

of 1.5 g cm-3 for uranium mill tailings or soils (RAC 1995, p. J-9).

Bulk density ρ of 1.5 g/cm3 is used for the K-65 material.

The decay constant λ of 222Rn is 2.1 × 10-6/s 

The radon emanation coefficient E is 0.28 (for oven-dried uranium mill tailings (Yu et al. 1993).

The 226Ra activity concentration R is 525 nCi/g in Silo 1 and 299 nCi/g in Silo 2 (RAC 1995, 

Table 8.1). 

The results of this method for the silos, substituting the values for R, E, λ, ω, and ρ in the above 

section into Equations 5, 6, and 7, yields a radon exhalation flux from the K-65 material of 7.10 × 10-7

Ci/m2

/s and 2.75 × 10-7 Ci/m2/s in Silos 1 and 2, respectively. The total rate of radon emanating from 

the K-65 raffinate in the silo airspace is then the product of the exhalation flux and the surface area of 

the raffinate. It was assumed that the dome approximates a circle with a 40-ft radius, and thus the 

surface area of the raffinate is 467 m2

. This resulted in 7,100 Ci/yr and 4,050 Ci/yr

2.3.2

of radon entering

into the headspace from the raffinate for Silos 1 and 2, respectively.

The calculation continues by providing analytical treatments for determining (1) the amount of radon 

released to the environment that was driven by the diurnal temperature differential and the Venturi 

effect of the prevailing wind speed, (2) the retention time of radon gas in the headspace affected by 

the containment structure of the silo, and (3) the depletion of 222Rn in the headspace due to the 

radioactive decay process before release to the environment.

Calculation for the Radon Rate of Release to the Atmosphere 

Once created in the K-65 materials, the radon would migrate to the headspace of the silo as 

described and quantified in Section 2.3.1. Being heavier than air, the radon would tend to stay close 

to the bottom, and would be moved toward the direction of its decreasing concentration in the dome 

airspace by atomic diffusion, unless it were mechanically stirred or pulled up by the upward air 

movements that were being driven to the openings at the top of the silo by the diurnal changes in 

temperature and the Venturi effects of the wind flow. These openings consist of a 6-in.-diameter 

gooseneck vent pipe and the gaps between the manholes and the manhole covers. Although RAC 

(1995) describes significant cracking in the walls that could provide flow paths for the migration of air 

to the outside, the extremely large resistance in the cracks would direct nearly all the airflow toward 

the unimpeded 6-in.-diameter vent pipe and the manhole cover gaps at the top of the dome. This 

conclusion is supported by the abrupt decrease in the environmental 222Rn concentrations after the 

sealing of the manhole covers and the removal of the gooseneck vent pipe in 1979. 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 14 of 76

The leak path factor (LPF) is conventionally used to predict the release of radioactive materials from 

their containment structures such as, in this case, the silos. The LPF is defined as the fraction of 

airborne radioactive materials that escape from their confinement structure via available pathways 

(i.e., the ratio of the radon released to the environment to the net radon source term in the headspace

of the silos). LPF calculations were made by developing and applying physically realistic modeling 

using the Nuclear Regulatory Commission’s generalized mass transport and thermal hydraulics 

computer program CONTAIN 2.0 (SNL 1997). Although this program was developed to predict the 

thermal hydraulic response inside containments and the release of radionuclides to the environment 

in the event of containment failure, the modeling capabilities of CONTAIN are sufficiently flexible that 

it can be applied to the analysis of containment structures in general.

The radon exhalation rate in Section 2.3.1.1 was used as the initial radon source term in the 

headspace. This amount is reduced by radioactive decay for the duration of the radon residence time 

in the headspace. The residence time is the period in which radon resides in the silo before exiting 

the silo. The residence time of the radon can be determined by dividing the flow rate of the radon-air 

mixture leaving the silo by the volume of the headspace. Subsequently, the amount of radon that had 

decayed during that period can be calculated. Only the surviving radon would reach the environment. 

As such, the net radon source term in the headspace of the K-65 silos (QRn) is represented by the 

following equations:

QQRRRR = qqRRRR (1 − ffdd) (8)

where

QRn = the net amount of 222Rn in the K-65 silo headspace.

qRn = the exhalation rate of 222Rn from the K-65 material

fd = the fraction of 222Rn depleted due to radioactive decay in the headspace of the silo.

The fraction of radon decayed fd is calculated by:

ffdd = 1 − ee−λλRRRR tt (9)

where

λRn = the decay constant of 222Rn = 0.00755/hr

t = the residence time of 222Rn in the headspace

The residence time t is obtained by:

tt = VVhss

qqeeee (10) 

where

Vhs = the volume of the silo headspace

qex = the exiting radon-air mixture flow rate from the silo due to diurnal and wind effects 

Substituting Equations 9 and 10 into Equation 8 yields Equation 11 showing the net 222Rn source term 

in the K-65 silo QRn as a function of the radon-air mixture exiting flow rate qex due to the diurnal 

temperature difference and the Venturi effect from the wind:
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 15 of 76

(11) 

2.3.2.1 CONTAIN 2.0 Leak Path Factor Calculations

The CONTAIN 2.0 computer program (SNL 1997) is an integrated analysis tool used for predicting the 

physical conditions, chemical compositions, and distributions of radiological materials inside the 

containment and connected buildings of a nuclear reactor after the release of material from the 

primary system in a light-water reactor accident. It is designed to predict the thermal-hydraulic 

response inside containments and the release of radionuclides to the environment. CONTAIN 

analyzes the thermal-hydraulic phenomena, aerosol behavior, and fission product processes, and the 

interactions among these phenomena. 

The code treats a containment system as a network of interconnected control volumes or “cells.” A 

compartment can also be partitioned to model phenomena such as natural convection and 

stratification within the compartment. The cells communicate with each other by means of mass flow 

of material between cells and/or heat conduction between cells through heat transfer structures. The 

flow directions through the flow paths are determined by the code calculation. Because there is 

considerable flexibility in specifying the properties of each cell and the connections between cells, the 

code is able to handle a wide variety of containment types.

Figure 1 defines the physical configurations of the silos for the CONTAIN code modeling. 

Head space air

qex3

26 ft-8 in

Earth Berm

Earth Berm

9 ft-4 in

Prevailing wind

Nominally 4 in concrete dome

80 ft

Unloading manhole Instrument manholes 

6 in diameter 

gooseneck vent

8 in concrete

wall

qex2

qex1

Radium containing K-65 material 

(Porosity from 0.6 to 0.8)

QRn

qRel

Figure 1. K-65 silos at the FMPC. 

The LPF is defined as:

LLLLLL = qqrr rr

QQRRRR (12)

where

qrel = the radon released to the environment
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 16 of 76

QRn = the net amount of 222Rn in the K-65 silo headspace [Equation 11]

and: 

qqeeee = qqeeee1 + qqeeee2 + qqeeee3 (13) 

where

qex1 = the radon-air mixture exiting flow rate from the gooseneck vent pipe

qex2 = the radon-air mixture exiting flow rate from the unloading manhole cover gap

qex3 = the radon-air mixture exiting flow rate from the four instrument manhole cover gaps

The following assumptions were applied to the development of the CONTAIN input parameters:

• Flow through wall cracks are small in comparison with flow through the 6-in. vent and manhole 

cover gaps

• Total volume of a silo is about 160,000 ft3 (4,500 m3

) (RAC 1995)

• The silo is two-thirds full of K-65 materials with a porosity of 0.7 (RAC 1995)

• The diurnal temperature difference is 19.46°F (RAC 1995)

• The prevailing wind speed is 6.5 mph (ORAUT 2006b) 

• Each instrument manhole has a diameter of 6 in. with 4-in.-thick concrete. The gap around the 

circumference of the manhole and manhole cover is 0.125 in.

• The unloading manhole has a diameter of 24 in. with 4-in.-thick concrete. The gap around the 

circumference of the manhole and manhole cover is 0.125 in.

• The total length of the 6-in. gooseneck vent is 5 ft. 

• CO2 gas is used as a surrogate for radon. 

The K-65 silo is modeled as a series of nine cells and the environment is represented by the 10th cell. 

Cell 1 contains the radium-bearing K-65 materials having 0.7 porosity. Its volume counts only the 

portion available to the gas. Cell 9 is the domed headspace and is simulated by a cylinder. The 

connections of Cell 9 to the outside environment (Cell 10) are by three engineered vents: one for the 

unloading manhole, one for the gooseneck vent pipe, and one for the four instrument manholes 

combined. Cells 2 through 8 represent the remainder of the silo and are for observing gradients in the 

gas. All internal cells contain air with various molar fractions of CO2. The CONTAIN program has the 

ability to track materials with thermodynamic properties if they are in the code library, such as CO2. At 

1 atmosphere and 300 K, CO2 is about 53% denser than air. Carbon dioxide gas works well as a 

surrogate for radon and is claimant-favorable because radon is significantly denser than CO2, which 

leads to an overestimation of releases to the environment. Cell 10, the environmental cell, initially 

contains only air and the arriving mass of CO2 over time is tracked by CONTAIN.

Attachment A presents the details of the CONTAIN calculations.
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 17 of 76

2.3.2.2 CONTAIN Code Results

Figure 2 shows the flow rates for the three vents. The total flow to the environment promptly reached 

a value of qex = 17.88 cfm. Figure 3 shows the CO2 mass in Cell 10 (environment) versus time for a 

5% CO2 molar fraction in the silo. The resulting CO2 mass in Cell 10 was divided by the total initial 

CO2 mass in the silo to obtain the values for the LPF. Figure 4 shows the LPF versus time for the 5% 

CO2 and air mixture in the silo. The LPF value reached a plateau of 0.873% after 4 hours of release. 

The LPF calculation was repeated with 10%, 15%, and 20% CO2 molar fractions and the 

corresponding LPFs were 0.864%, 0.855%, and 0.847%. The LPFs changed only slightly, decreasing 

as the molar fraction increased. The results indicate that the fraction vented is not sensitive to the 

molar fraction of CO2 used. Therefore, it can be concluded that a value of 0.873% is an appropriate 

LPF (the fraction of radon vented to the environment) regardless of the amount of radon in the 

headspace for this problem.

Figure 2. Flows to the environment from the engineered vents. 

2.3.3

The radon release from the K-65 silos can be obtained by solving Equation 14:

Environmental Radon Release Rate

qqrr rr = LLLLLL × QQRRRR (14)

where

LPF = the leak path factor, 0.00873

QRn = the net source term in the headspace 

= 

20

18

16

14

12

10

8

6

4

2

0

Flow to Environment (cfm)

0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 1400 1500

Time (sec)

Unload Manhole Flow

Gooseneck Flow

Instrument Manhole Flow

Total Flow

 Problem M2-D

All silo cell atmospheres contain

 5% CO2 by molar fraction
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 18 of 76

3.0

2.5

2.0

1.5

1.0

0.5

0.0

Mass of CO2 Vented to Environment (kg)

14x1030 2 4 6 8 10 12

Time (Sec)

 Problem M2-D

All silo cell atmospheres contain

 5% CO2 by molar fraction

Total initial CO2 mass in silo = 339.9563 kg

Initial CO2 mass in cell 1 = 159.074 kg

Initial CO2 mass in cell 9 = 63.2576 kg

Initial CO2 mass in cells 2-8 = 113.6247 kg

Figure 3. Flows to the environment from the engineered vents. 

1.0

0.8

0.6

0.4

0.2

0.0

% CO2 Vented to the Environment

14x1030 2 4 6 8 10 12

Time (Sec)

 Problem M2-D

All silo cell atmospheres contain

 5% CO2 by molar fraction

Figure 4. Percent of the CO2 vented to the environment versus time. 

 where:

qRn = 7,100 Ci/yr Silo 1 and 4,050 Ci/yr Silo 2

λRn = 0.00755/hr

Vhs = 4.00 × 104 ft3 (RAC 1995, p. J-30)

qex = 17.88 × 60 = 1,073 ft3/hr (From Figure 2)

Solving Equation 14 yields the combined (Silos 1 and 2) annual environmental releases before 1980 

of 70.4 Ci/yr. 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 19 of 76

2.3.4

The alternative calculation of the radon release from the K-65 silos indicates a release rate of 

70.4 Ci/yr, which is slightly higher than the 60-Ci/yr diffusion release calculations performed by Borak 

(1985). The incremental increase in release rate primarily represents the amount of radon that 

escaped through the openings on the top of the dome driven by the daily temperature difference and 

the Venturi effect from the wind. All previous calculations assumed a constant radon concentration in 

the silo air mixture. Being a mass transport and thermal hydraulics computer code, the air/radon air 

mixture flow rates through the dome openings to the outside environment predicted by CONTAIN 

based on physical configurations and thermal-hydraulic parameters are particularly accurate. Figure

2 shows a total flow to the environment from the engineered vents of approximately 18 cfm. When 

using the actual radon concentration in the headspace as shown in RAC (1995, Table J-12) (2.5 × 107

pCi/L for Silo 1), the alternative calculation would have resulted in a radon release rate to the 

atmosphere of 

Summary Conclusion

6,698 Ci/yr: 

2.5 × 107pCi/L × 2.83 × 10−11 Ci/ft3 × 18 ft3/min × 5.26 × 105min/yr = 6,698 Ci/yr

The alternative result of 6,700 Ci/yr is practically identical to the RAC (1995) results of 5,000 to 

6,000 Ci/yr for the same period, and both calculations are in line with each other. On the other hand, 

perhaps the RAC report could be proven to have overestimated the radon release by CONTAIN.

Because of the very large uncertainties associated with the SC&A 210Pb deficit radon release 

calculation and because its validity has not been verified or reviewed by the National Academy of 

Sciences, 

2.4 RESPONSE TO APRIL 2010 SC&A WHITE PAPER

NIOSH will continue to adopt the claimant-favorable model for radon releases, as given in 

Section 5.2.4 of ORAUT-TKBS-0017-5 (ORAUT 2004), that 5,000 to 6,000 Ci/year of 222Rn were 

released from the silos.

The purpose of this section is to respond to the salient points in the Second White Paper (SC&A 

2010) in relation to the NIOSH dose reconstruction position to:

....rely upon the individual exposure estimates produced using the Pinney/Hornung 

model which utilized radon exposure levels from the RAC study, plus an additional 

radon source term at Fernald. ...

The primary negative conclusions drawn by SC&A regarding the Pinney Report (Hornung 2008) are:

(1) The Pinney Report simply accepted the source term releases of radon from the K65 Silos, as described by the following two RAC reports issued in 1995 and 

1998:...

(2) Using the modeled RAC (1995)/RAC (1998) radon release source terms, the 

“Pinney Report” investigators applied a modified Gaussian dispersion model 

originally ...

These conclusions failed to recognize the extensive efforts to validate the RAC model by the 

Hornung/Pinney team. The description of the validation of model estimates in Hornung (2008) is 

repeated here to ensure their efforts are not overlooked.

From Pages 3 and 4 of the Pinney Report (Hornung et al. 2008):
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 20 of 76

Validation of model estimates

The original Gaussian plume model was calibrated to outdoor measurements of radon 

made at various locations on the site during the years 1981 through 1987 (Killough and 

Schmidt, 2000), with estimated emission rates from the K-65 silos as input. Before 

1979, when vent pipes were removed and the silo domes were capped, very few radon 

measurements were available. The RAC dose-reconstruction team made a thorough 

search for any historical measurements that could have been used to calibrate their 

model, but the era of highest exposures (1952–1979) was difficult to validate. We 

sought a way of validating our exposure estimates, particularly before 1979. Several 

case–control studies of radon related lung cancer from exposures received in homes 

(Field, 2001; Steck et al., 2002) employed a method for estimating cumulative 

exposure. This method involved the use of CR-39 plastic film applied to glass in 

mirrors or framed pictures that were kept in each home that a case or control inhabited 

over their adult lifetime. The CR-39 film records tracks from 210Po a-particles that 

decay from 210Pb which is a long-lived (22 years half-life) decay product of 222Rn 

imbedded in the glass. Counts of tracks from the CR-39 plastics can be used to 

estimate integrated radon exposure over long periods of time (Lively and Steck, 1993; 

Mahaffey et al., 1993). 

Since CR-39 had been primarily used in laboratories and homes, a test of its feasibility 

in a uranium processing facility was necessary. Eight CR-39 films were obtained from 

Track Analysis Systems Ltd. (TASL) in Bristol, England. They were placed on the 

inside and outside of glass window panes in a building near the center of the 

production area and in a building near the eastern edge of the plant farther from the K65 silos. Four replicates were made by placing films side-by- side on the same pane of 

glass. These films were exposed for approximately 2 weeks in November 1999 and 

then sealed and sent to TASL for analysis with no indication of where they had been 

placed. Upon receipt of the exposed CR-39 films, TASL etched the films in a solution 

of NaOH, calibrated them with a 252Cf fission source, and scanned them with an 

automated image analysis system. Surface activity of 210Po is measured with 

corrections for lower energy particles from 234U, 235U, and 238U and similar lower energy 

α particles (Fews and Henshaw, 1982).

Higher energy contamination (214Po, 218Po, and 212Po) was eliminated from track counts 

and did not cross contaminate 210Po readings. The actual measurements were made 

by etching and scanning TASTRAK α-particle sensitive CR-39 plastic, followed by 

spectroscopic analysis of associated a-tracks (Fews, 1992). When track density was 

extremely high, tracks were counted by eye using averages of 10 microscope image 

frames randomly distributed over the sample. Results for all samples were reported as 

total 210Po activity per unit area in Bq/m2

. 

In addition to the use of CR-39 film for validation of the air dispersion model, a set of 

radon measurements were discovered that were not used to validate the model when it 

was developed to estimate off-site radon levels for residents near Fernald. These data 

were collected for 7 months from March to September 1991 by a student for a Master’s 

degree thesis (Cardarelli, 1992). Samples were taken around the K- 65 silos and in the 

vicinity of five buildings on the Fernald site. The protocol for the Master’s thesis called 

for a survey using three monitors simultaneously: Pylon monitors, Radon Gas 

monitors (RGMII), and Femto-tech monitors, all three located at each measurement 

site. The Pylon monitor utilizes passive diffusion of radon gas into a scintillation 

chamber where a-particles are counted. The RGM II utilizes a pump to draw air 

through a particulate filter into a scintillation chamber. The Femto-tech monitor is 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 21 of 76

similar to the Pylon monitor in using passive diffusion, but it employs a pulsed ionsensing chamber instead of a scintillation chamber. The data available for our 

validation consisted of the arithmetic mean (AM) of the three monitor results for each 

hour. The hourly means were then used to calculate geometric means (GMs) for each 

8-h shift across the 7 months when sampling was done since the distribution of radon 

levels over time was right-skewed.”

The Pinney Report (Hornung et al. 2008) was not the first time the validation of the RAC model was 

described. The validation was also described in Pinney et al. (2004) by the Hornung/Pinney team on 

their work to assess the radon and cigarette smoking exposure of Fernald workers. Again the radon 

dispersion model developed by RAC for emissions from the K-65 silos was used and validated in the 

assessment. 

From page 5 of Pinney et al. (2004):

Since the RAC model is a conceptual mathematical model developed from physical 

principles and is not based directly upon measured data, we attempted two ways to 

investigate the accuracy of its estimates. First, we discovered radon measurement 

data that had not been used by the RAC group to calibrate their model. These were 

data collected by John Cardarelli for his Master's thesis during the period of March 

through September 1991. Data were collected both in the immediate vicinity of the K65 silos and at various points around the production area. We then compared 

Cardarelli's measurements by shift and location to the closest year of our estimates, 

1988. Since no remediation was done between 1988 and 1991, we assumed that 

results should be comparable. Results are shown in the attached Tables 2 and 3.

Table 2. Comparison of model predictions and 1991 radon 

measurements.
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 22 of 76

Table 3. Comparison of model predictions to 1991 silo 

measurements.

Our second attempt to validate the model was the use of CR-39 film attached to 

windows in selected buildings around the site. The CR-39 film measures alpha activity 

from the decay of Pb-210, which was embedded in the glass over several decades. 

This assay, therefore, serves as a cumulative measure of radon and its decay products 

since each building was constructed. .... With NIOSH support, we expanded our 

sample size to 110 samples placed throughout the site on buildings selected for their 

proximity to both the K-65 silos and the plant 2/3 area where we found the initial high 

measurements. Results of the larger survey indicated that glass sampled near the 

Q-11 silos in the plant 2/3 area were substantially higher than glass closer to the K-65 

silos. The Q-11 silos had been used in the period 1953-58 to store highly radioactive 

ore before it was processed in the plant 2/3 complex. .... Once we became aware of 

these CR-39 findings, we conducted additional document searches and met with 

workers several times to develop a better understanding of this local source of radon.

2.5 CONCLUSION

The Hornung/Pinney team validated the RAC model estimates with an independent set of radon 

measurement data that had not been used by the RAC group to calibrate their model. These data 

were collected by John Cardarelli for his Master's thesis from March through September 1991. Data 

were collected in the immediate vicinity of the K-65 silos and at various points around the production 

area. Tables 2 and 3 of Pinney et al. (2004) show that the comparisons were in good agreement with 

the Killough model predictions for the first and second shifts. The model predicted higher exposures 

that were claimant-favorable. It also predicted lower exposures for the third shift and the differences, 

in general, are within a factor of 2. In the scientific community, a factor-of-2 difference is a reasonable 

validation of the model predictions for this type of analysis. 

NRC (1994, p. 14) states: 

Measurements were used for comparison with the estimates derived from the model, 

and the overall results are verified to be within a factor of 2. ... Overall, the results of 

the comparison are a reasonable validation of the model predictions.

The Pinney/Hornung model on which NIOSH will rely for individual exposure estimates is entirely 

credible and appropriate because the analytical model had initially been calibrated with measurement 

data and later validated with independent sets of measurement data. 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 23 of 76

3.0

3.1 PROCESSING HISTORY OF PITCHBLENDE AND DOMESTIC URANIUM FEEDSTOCK

Q-11, K-65, DOMESTIC URANIUM, AND RAFFINATES

During the Manhattan Project, extremely high-grade (50% to 80%) uranium ore was imported from 

African Metal Company (AMC) in the Belgian Congo and was designated with the code “Q-11.” At 

this level of uranium content, the ore could be processed directly in the modified plutonium-uranium 

extraction (PUREX) process at FMPC. From 1946 until 1953, Mallinckrodt Chemical Works (MCW) in 

St. Louis, Missouri, produced uranium from the Q-11 ore. The Q-11 ore contained all of the uranium 

daughter products, including 226Ra. When the ore was refined, 226Ra and other daughter products 

were concentrated in the waste stream, which was identified with the code “K-65.” The term raffinate

was used to describe the waste stream. The K-65 residues were returned to Belgium until March 

1949 (NLO 1979; Heatherton 1949). In 1949, an agreement was reached wherein the U.S. Atomic 

Energy Commission (AEC) would store the K-65 residues for later return to AMC. Beginning on 

March 22, 1949, the K-65 residues from MCW were sent to Lake Ontario Ordnance Works (LOOW) to 

be stored (Heatherton 1949). 

Construction of FMPC began in 1951 and MCW began shipping K-65 material to FMPC rather than

LOOW (Walden 1952; Vogel 1989). Between September 25, 1951 and April 30, 1952, 11,253 drums 

were sent from MCW to FMPC; a total of 25,000 drums were shipped in all (Vogel 1989). The 

shipments of K-65 residue from MCW continued until January 1956 (Vogel 1989).

When construction was completed, storage of all K-65 material was assigned to FMPC and two large 

ground-level silos were built for this purpose. In addition to the residues shipped from MCW and 

LOOW, FMPC began processing Q-11 ore and generating its own raffinate streams that needed to be 

stored. Pitchblende ore processing resulted in “hot” raffinates, indicating that relatively high 

concentrations of 226Ra plus daughter products were in the waste stream with the associated large 

increase in external radiation fields. The transfer of K-65 material to the silos began in late 1952 

(Wunder 1953a). The silos were designed to hold about 24,000 drums each (Vogel 1989). Silo 1 

contained K-65 material from LOOW and MCW. Silo 2 held K-65 material generated at FMPC and 

from off site, including MCW (Vogel 1989). 

The transfer of the drummed K-65 (the hot raffinate) material to the silos took place at a drum 

handling station, where the material was dumped from the drum to a slurry tank. The slurried material 

was pumped to the storage silo and the excess water returned to the slurry tank. This operation did 

not run continuously; for example, freezing weather caused the dumping to be stopped on January 9, 

1953, for about 2 months. At that time, Silo 1 contained 20,259 drums, with 2,257 drums from LOOW 

remaining (Wunder 1953a). In June 1953, the K-65 dumping station was being operated on the day 

shift only, the dumping was current on MCW material, and there were 671 drums remaining from 

LOOW. The south tank (Silo 1) was reported to be full and the dumping was moving to the north tank 

(Silo 2) (Wunder 1953b). The last of the K-65 material from MCW was dumped in April 1956 (Vogel 

1989).

K-65 material from the processing of Q-11 ores at FMPC was initially transferred in 1954 and 

periodically through 1957 (Vogel 1989, Attachment 8). This processing occurred in Plants 2 and 3. 

See Section 3.4 for specific dates of these campaigns and a listing of other K-65 materials that were 

transferred to Silo 2. The total K-65 material in the two silos was 19,385,126 lb, including about 

380,000 lb from the processing of “rum jungle” ore from Australia in 1957 and 1958 (Vogel 1989,

Attachment 18; Shaw 1968; Robinson 1973). About 25% of the material in the two silos was 

generated at FMPC (Robinson 1973). In summary, the following events or campaigns contributed to 

the history of the K-65 silos:

1946–1953 All uranium ore processing occurred at MCW. 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 24 of 76

1946–1949 K-65 residues were sent from MCW to Belgium. 

3/1949–1951 K-65 residues were no longer sent to Belgium, but were sent from MCW to LOOW.

1951–1/1956 MCW shipped K-65 residues directly to FMPC. 

9-11/1952 Drums of MCW residue that were not dumped into the LOOW water tower were 

shipped to FMPC under the supervision of FMPC, but using workers from the 

Niagara Falls area.

Late 1952 FMPC began dumping drums of K-65 from MCW and LOOW to Silo 1. 

June 1953 The drum dumping operation was current on MCW drums, with 671 drums left from 

LOOW; the south tank was full (Wunder 1953b). 

4/54–4/57 Processing of Q-11 ores at FMPC occurred periodically during this period with the 

associated transfers of K-65 materials from the local process.

4/1956 The last K-65 material from MCW was transferred to Silo 2.

1957–1958 Australian pitchblende was processed with the raffinates being transferred to Silo 2.

Most U.S. and Canadian ores were low grade (1% or less uranium content) and required processing 

near the mine sites. The product from these mills was a “yellowcake” (U3O8) of approximately 60% 

uranium content. After 1958, when processing of the Q-11 and other pitchblende ores at FMPC had 

been completed, the source of the uranium for processing at FMPC was this yellowcake. The 

yellowcake had most of the radium and other impurities removed before being shipped to FMPC. 

This raffinate was called “cold raffinate” because the waste stream had much lower concentrations of 

226Ra plus daughters, compared with the K-65 raffinate. The cold raffinate was a calcined dry, 

dispersible powder that was transferred to Silo 3 using an enclosed air lift. (The fourth silo was never 

used and remained empty.)

The isotopic contents of the three silos are shown in Table 1. 

Table 1. Isotopic composition in FMPC Silos 1, 2, and 3. 

Isotope

Silo 1

activity 

concentration 

(nCi/g raffinate)

Silo 1

percent 

activity

Silo 2

activity 

concentration 

(nCi/g raffinate)

Silo 2

percent 

activity

Silo 3

activity 

concentration 

(nCi/g raffinate)

Silo 3

percent 

activity

Uranium 1.68 0.2 2.37 0.3 3.63 4.8

Ac-227 7.67 0.8 6.64 0.8 0.93 1.2

Pa-231 N/A 0.0 4.04 0.5 0.63 0.8

Pb-210 202 20.0 190 24.3 3.48 4.6

Po-210 281 27.8 231 29.6 N/A 0.0

Ra-224 N/A 0.0 N/A 0.0 0.37 0.5

Ra-226 447 44.2 263 33.6 3.87 5.2

Ra-228 2.28 0.2 N/A 0.0 0.41 0.5

Th-228 N/A 0.0 7.36 0.9 0.75 1.0

Th-230 68.9 6.8 76.2 9.7 60.2 80.1

Th-232 1.11 0.1 0.99 0.1 0.84 1.1

Total 1011.64 N/Aa 781.6 N/A 75.11 N/A

a. N/A = not applicable.
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 25 of 76

Several observations are made from these data. Silos 1 and 2 contain the K-65 (hot) raffinates and 

include relatively high concentrations of 226Ra. Silo 3 contains cold raffinate with very little 226Ra. 

Because uranium is the extraction product, the raffinates in the silos contain low levels of uranium. 

Thorium-230 is present in each silo in similar concentrations, indicating that the uranium mills 

effectively removed radium but were not effective in removing thorium from the yellowcake. The total 

activity concentration in Silo 3 is approximately 10% of the activity concentration in Silos 1 and 2.

These observations indicate potential problems for dose reconstruction techniques that are based on 

uranium bioassay data because the uranium intake when handling these materials might not have

been the dominant isotopes. As a consequence, alternative methods of dose reconstruction are 

defined.

3.2 AVAILABLE MONITORING DATA 

An extensive air sampling program was conducted at FMPC. The program was designed for worker 

exposure control by monitoring radioactive air dust to identify process areas that needed better 

engineering controls or designs, and to identify operations that required respiratory protection. These 

studies involved air samples in all the production plants, and included general area samples and 

breathing zone samples for workers. The studies were documented in annual exposure studies from 

1955 to 1968 (AEC 1955; Stefenac 1955a,b; Stefenac and Schumann 1956a,b; Halcomb and 

Huesing 1957; NLO 1957; Wing and Halcomb 1958a,b, 1959a,b; Wing and Ruhe 1960a,b; Wing et al. 

1961; Wing, Ross, and Cline 1961; Wing, Ross, and Leininger 1962; Wing and Leininger 1965; Wing, 

Leininger, and Ruhe 1966; Ross, Leininger, and Lawrence 1968; Ross, Leininger, and Zimber 1969).

Tables 2, 3, and 4 are summaries of the results from these exposure studies in Plants 2 and 3, where 

feed materials were charged to the processes and raffinate processing occurred, for each job 

category and each year. The units are in multiples of the maximum allowable concentration (MAC). 

See Section 2 for more information about DWE reports.

Several observations can be made on plant operations and personnel exposures from these data.

• Data were collected separately for Plants 2 and 3 from 1955 to 1962. The two plants were 

shut down for the next 2 years. After restart, data were collected as combined Plant 2/3 data. 

• The DWE for each job category did not take into consideration that respirators were used by 

the workers for jobs that created high dust levels. As a consequence, actual exposures to the 

workers would have been lower than depicted in these tables.

• The data for 1955 through 1957 covered the end of the processing of the pitchblende ores 

from the Belgian Congo and Australian sources in 1958 with high radium content. The 

raffinate stream from these ores was stored in Silo 2. The processing then switched to 

yellowcake from mills in the United States and Canada, which continued until Plants 2 and 3 

were shut down in 1962. It was during this period that the raffinate stream was sent to Silo 3 

for storage.

• The level of exposures to air activity as measured by the DWE sampling program gradually 

decreased over the 14 years of operations. The exposure studies attribute the decreases in 

part to improvements in process controls and procedures.

• The studies for Plants 2 and 3 indicate that the highest airborne dust concentrations were 

measured in the head end (ore and concentrate) dumping operations. The ore dumping 

operations were dusty and less well contained. The high values for the ore dumping operators 

were actually the results of very high values experienced while dumping drums containing the 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 26 of 57

Table 2. DWE data for Plant 2, 1955 to 1962, when Silos 2 and 3 were being filled. The # field represents the number of workers 

potentially exposed at that daily weighted value at the time the data were collected and analyzed by the FMPC staff.

Job category

1955a 1956 1957 1958 1959 1960 1961 1962

# DWE # DWE # DWE # DWE # DWE # DWE # DWE # DWE

Ore dumping and metal 

dissolver operators

3 2.81 3 0.76 6 4.1 3 8.3 3 24

Ore dumping 3 0.9 4 6.2 4 2.16 6 4.26 6 5.9 3 1.4

Extraction operators and 

helpers

24 0.21 15 0.53

Digestion operators and 

helpers

24 0.64

Laborer 3 1 12 0.87 12 0.76

Boildown and denitration 

operators

24 0.74

Gulping helper 6 0.85 4 0.76

Gulping operator 6 0.05 12 0.82 8 0.15

Leadman 1 0.49 1 0.16

Control room 6 0.06 4 0.73 4 0.16

Pot man 8 0.31 8 0.15 6 0.96 3 0.71 5 0.1

Floor man 3 0.37 8 0.19 3 0.81 6 0.21 3 0.56 3 0.1

Utility operator and helpers 3 0.06 12 0.32 12 0.19 3 0.81 3 0.25 3 0.54 3 0.1 3 0.2

Forklift operator 3 0.46 3 0.15 3 0.56 3 0.1 3 0.2

Shift clerk 4 0.35 4 0.11 2 0.46 2 0.13 2 0.49 2 0.1 1 0.2

Leadman, operators, and 

helpers

27 0.65 30 0.21 20 0.11

Evaporator operator 4 0.31 4 0.13 3 0.63 3 1.24 3 0.9 3 0.2 2 0.2

Area foreman 1 0.06 1 0.3 1 0.11 1 0.57 1 0.26 1 0.42 1 0.1 1 0.2

Shift foreman 3 0.06 4 0.26 4 0.12 3 0.79 6 0.34 3 0.39 4 0.1 4 0.2

Pump man 3 0.57 3 0.31 2 0.23 2 0.1

Production records clerk 5 0.12 3 0.16 3 0.12 2 0.23 2 0.1 2 0.2

General foreman 1 0.12 1 0.2 1 0.12 1 0.53 1 0.35 1 0.3 1 0.1 1 0.1

Analyst 3 0.25 4 0.73 4 0.08 3 0.1 3 0.24 3 0.22 3 0.1 3 0.2

Extraction operator 24 0.21 18 0.37 15 0.2 9 0.1

Superintendent 1 0.11 1 0.2 1 0.06 1 0.37 1 0.22 1 0.17 1 0.1 1 0.2

Office assistant; admin. 

assistant

1 0.11 1 0.21 1 0.12 2 0.14 2 0.11 2 0.14 1 0.1 1 0.2

Sump operator 6 0.09 8 0.18 8 0.05 6 0.09 6 0.2 3 0.05 3 0.1 3 0.1

Solvent reclaimer 6 0.1

Compiled from two exposure studies (Stefanec 1955a, Stefanec 1955b) performed in 1955. 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 27 of 57

Table 3. DWE data for Plant 3, 1955 to 1962, when Silos 2 and 3 were being filled. The # field represents the number of workers 

potentially exposed at that daily weighted value at the time the data were collected and analyzed by the FMPC staff.

Job category

1955 1956 1957 1958 1959 1960 1961 1962

# DWE # DWE # DWE # DWE # DWE # DWE # DWE # DWE

Pot man 6 0.63 6 0.55 6 0.8 4 0.2

Leadman 6 0.44 6 0.23 6 0.09

Laborers 3 1 7 0.21 7 0.08

Gulping operator 6 0.09 6 0.73 6 0.46 6 0.6 5 0.2

Utility operator 3 0.5 1 0.2

Area foreman 1 0.25 1 0.07 1 0.18 0.18 1 0.2 1 0.1 1 0.2

Shift foreman 4 0.53 4 0.23 4 0.09 3 0.37 3 0.33 3 0.5 8 0.1 1 0.1

Systems operator 3 3.77 3 0.4

Recovery operators and 

helpers

24 0.51 29 0.22 27 0.36 12 0.14 12 0.08 12 0.1 5 0.1 5 0.1
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 28 of 76

Table 4. DWE data for Plant 2 and 3, 1965 to 1968, during low 

acid processing period. The # field represents the number of 

workers potentially exposed at that daily weighted value at the time 

the data were collected and analyzed by the FMPC staff.a

Job category

1965 1966 1967 1968

# DWE # DWE # DWE # DWE

Digestion operator 4 1.7 6 1.6 12 4.2 3 2.9

Denitration operators 6 0.3 9 2.2 6 0.5 4 1.7

Foreman 3 0.2 3 0.2 3 0.3 2 0.7

Clerk 1 0.2 1 0.1 1 0.1 1 0.1

Superintendent 1 0.2 1 0.2 1 0.2 1 0.2

Hot raffinate 3 0.1 3 0.1 3 0.2 3 0.2

Combined raffinate 4 0.1 5 0.1 3 0.1 3 0.2

Forklift operator 1 0.1 1 0.2 3 0.5 3 0.5

Shift foreman 6 0.4

Prod. records clerk 2 0.1

Extraction operator 6 0.1 6 0.1 3 0.1 3 0.1

Sump operator 3 0.1 3 0.1

a. The low acid period refers to a change in process from concentrated nitric 

acid (which for economic considerations required recovery by condensation 

from a calciner) to a more dilute acid process, which could be neutralized and 

pumped directly to waste disposal Pit 5.

uranium feed material into the head end of the process that were weighted or averaged with 

the time that was spent in other activities performed in lower air dust concentrations.

• The lowest concentrations were experienced by the workers in raffinate operations. The 

solvent extraction process was a liquid process completely contained in a closed tank/piping 

system. The workers in the raffinate areas experienced airborne concentrations that were 

consistently at the same levels as those assigned to records clerks and administrative 

assistants. These levels should be considered ambient levels very near the detection limit of 

the monitoring equipment and techniques. The value of 0.1 MAC is the same value 

consistently reported for other nonproduction plant areas.

The most significant exposure potential from the “hot” raffinates was the relatively high 

external exposure rates due to high concentrations of 226Ra and daughter products. 

Shielding was provided for the equipment that was a part of the Q-11 processing to minimize 

external exposures to the workers. The handling of K-65 drums from MCW and LOOW was 

recognized as having high external exposures; workers were rotated through this work activity 

to meet the external exposure guidelines (Heatherton 1952a).

• The DWE data for Plant 2/3 during the period from 1965 through 1968 clearly show the 

significant differences in the front end processes in Plant 2 compared to the raffinate 

processes in Plant 3. The raffinate operators experienced DWE levels that were 10 to 20 

times lower than the operators on the digestion/denitration side of the refinery.

These data demonstrate that exposures from the raffinate processes in Plant 3 were extremely low, 

essentially at background levels. The primary source of exposure for workers in Plants 2 and 3 was 

the beginning of the processing operation, the ore and concentrate charging operations. The source 

of the exposure is significant in establishing the ratio of the trace contaminants to which workers 

would have been exposed. The high thorium ratio observed in Silo 3 materials is associated with the 

liquid stream contained in a closed piping system and then to a closed calcination system, which 

provides little exposure potential to the workers. The process charging operation, which is the primary 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 29 of 76

source of exposure, has ratios of thorium to uranium that are typical of ore materials. At most the 

230Th would be in effective equilibrium with the 238U/234U activities in the feed material.

The 230Th concentration in Silo 3 is roughly the same (possibly 10% less) as the 230Th concentration in 

Silos 1 and 2. The 230Th in Silo 3 constitutes a high fraction of the total activity, compared with the 

fractions in Silos 1 and 2, due not to high 230Th levels, but solely to the absence of radium and its 

daughter products. If the 226Ra and its daughters were removed from Silos 1 and 2, the 230Th would 

constitute about 80% of the remaining relative hazard, as it does in the Silo 3 material. The 

concentration of 230Th is essentially the same in all three silos, indicating that the uranium milling 

processes removed little of the thorium from the ores. 

A review of the job categories in Tables 2, 3, and 4 indicates that the workers with the highest 

exposures in Plants 2 and 3 were the operators in Plant 2. Job categories that might have been 

considered transient, such as helpers, laborers, forklift operator, etc., are shown to have low 

exposures. In fact, helpers are included in the operator category in Plant 3, which experienced very 

low exposures. Any special maintenance activities that might have required work on the raffinate 

process systems would have been performed with appropriate protective equipment (respirators and 

ventilation), but primarily would have constituted a relatively short exposure time compared with the 

full-time exposure considered in the DWE exposure studies. As a consequence, such exposures 

would have been small in comparison. Therefore, the default recommendation to add an amount of 

230Th activity equal to the uranium intake activity is appropriate and conservative.

Radon Breath Analyses for Offsite Raffinate Dumping Operations to FMPC Sites

During 1952 through 1954, to infer radium intakes, 609 breath samples were taken for radon at the 

site. The samples were sent for analysis to the AEC Health and Safety Laboratory (HASL; now the 

Environmental Measurements Laboratory) in New York City. At this time, no specific information is 

available about the measurement method; the assumption is that the ambient radon concentration 

present during the sampling was included in the result. Therefore, any dose reconstruction based on 

these uncorrected results would include a claimant-favorable bias. 

A typical radon breath analysis result identifies the individual and provides the radon breath 

concentration measurement results in curies per liter. Annual and composite parametric values for 

the data are shown in Table 5. In all years, the minimum reported value is 1 × 10-13 Ci/L, which is 

assumed to be the reporting limit for the analysis. When the reported result is “<1 × 10-13 Ci/L” a 

value half of that (5 × 10-14 Ci/L) was used in calculations. This method is endorsed in NIOSH (1993), 

which is included by reference in the implementing rules for the Energy Employees Occupational 

Illness Compensation Program Act of 2000 (EEOICPA) [42 CFR Part 82.16(a)]. Approximately 25% 

of the values were imputed by this method. In addition, approximately 26% of the submitted samples 

were not successfully analyzed due to processing errors or equipment failures. Lacking data, these 

were removed from consideration, resulting in a total of 449 valid samples over the period from 1952 

through 1954. 

Table 5. Radon breath analysis statistical parameters. 

Parameter 1952 1953 1954

Composite 1952 

through 1954

Number of attempted measurements 140 238 231 609

N, number of valid measurements 84 183 182 449

Number of valid measurements less than 

the reporting limit

55 54 46 155

Minimum, Ci/L (assumed to be the 

reporting limit)

5.00E-14 5.00E-14 5.00E-14 5.00E-14

Maximum, Ci/L 9.00E-13 3.30E-12 1.20E-12 3.30E-12
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 30 of 76

Parameter 1952 1953 1954

Composite 1952 

through 1954

Mean, Ci/L 2.66E-13 2.84E-13 2.47E-13 2.66E-13

Std deviation 2.01E-13 3.26E-13 1.79E-13 2.53E-13

Coefficient of variation (std. dev./mean) 0.76 1.14 0.72 0.95

Geometric mean, Ci/L 2.01E-13 2.07E-13 1.91E-13 1.99E-13

GSD 2.18 2.13 2.11 2.13

Median, Ci/L 2.00E-13 2.00E-13 2.00E-13 2.00E-13

84th percentile, Ci/L 4.00E-13 4.00E-13 4.00E-13 4.00E-13

95th percentile, Ci/L 6.85E-13 6.90E-13 6.00E-13 6.00E-13

ORAUT-OTIB-0025, Estimation of Radium-226 Activity in the Body from Breath Radon-222 

Measurements, provides information on partitioning of the whole-body radium activity among bone, 

lung, and soft tissue compartments (ORAUT 2005). The radiation dose to various organs from 226Ra

in the body can be derived by selecting an appropriate intake scenario and applying appropriate dose 

factors to the derived 226Ra activity level. ORAUT-OTIB-0025 defines a whole-body 226Ra activity 

conversion factor for radon breath data as 2.52 × 105 pCi of 226Ra per pCi/L of exhaled 222Rn. 

Table 6 shows the ratio of three thorium isotopes found in the Silo 1 and 2 materials to 226Ra. These 

ratios are based on data in ORAUT-TKBS-0017-5, Technical Basis Document for the Fernald 

Environmental Management Project (FEMP) – Occupational Internal Dose (ORAUT 2004, Table 

5-16). The ratios from both silos are similar but the more claimant-favorable ratio (the higher value) 

from Silo 2 data is recommended for dose reconstruction use, because it is not possible to know 

which material was included with the radium intake.

Table 6. Ratios of Silo 1 and Silo 2 thorium isotopes to 226Ra. 

Isotope

Activity ratio of isotope to 

Ra-226 in Silo 1 material

Activity ratio of isotope to

Ra-226 in Silo 2 material

Th-228 0.0048 0.028

Th-230 0.14 0.29

Th-232 0.0023 0.0038

A comparison was made between the bounding scenario described in Fernald site profile (ORAUT 

2004, Section 5.2.4) and the bounding estimate used for this report. The bounding estimate for this 

report considers the 95th percentile estimate for 226Ra in the whole body of workers participating in 

the Silo 1 and 2 transfers. This bounding scenario was based upon air sampling taken during the 

drum transfer operations and the following assumptions: 

• Occupancy time limited by external dose (6 wk/yr),

• Alpha airborne activity at 100 MAC (4.5 × 10-9 μCi/mL) – maximum air activities measured,

• Breathing rate of 9.6 × 106

 cm3/d, 

• No use of respiratory protection equipment, and

• 46% of activity in the air sample being 226Ra (ORAUT 2004, Table 5-16, Silo 1).

The 95th-percentile whole-body estimate developed from the radon breath analysis is 0.15 μCi of 

226Ra. The whole-body estimate developed in ORAUT (2004) is 0.7 μCi of 226Ra. These estimates 

are reasonably close considering the uncertainty and in the ORAUT (2004) air sample-based 

bounding assumptions. The bounding estimate in ORAUT (2004) is favorable to the claimant. Based 

on the data in this report, it a lower the bounding estimate in the next revision of ORAUT (2004) could 

be justified. 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 31 of 76

3.3 RECOMMENDATIONS

Use radon breath analysis data to estimate the 226Ra whole-body burden for workers associated with 

K-65 raffinate handling or processing of offsite raffinate transfer to the K-65 silos. The 95th-percentile 

value observed in K-65 raffinate workers was 6 × 10-13 Ci/L of exhaled 222Rn, which according to 

ORAUT-OTIB-0025 is equivalent to 0.15 μCi of 226Ra (ORAUT 2005). This value should be assigned 

to K-65 raffinate workers.

Air monitoring results show that no significant exposure to radon from cold raffinates occurred. 

Unmonitored exposure to the impurities in uranium feed materials such as pitchblende and 

yellowcake might have occurred to workers who sampled or handled these materials. Workers in 

Plants 1, 2, 3, and 8 might have been exposed. The intake rates from these exposures can be 

bounded by adding an intake in the ore or concentrate in proportion to the uranium intake as 

measured by bioassay. Table 7 shows the recommended values for unmonitored isotopes that might 

be present in uranium mill concentrates. For dose reconstruction, the amount of uranium intake 

should be calculated from urine bioassay data. The results in mass quantities should be multiplied by 

the conversion factors in the table to derive the maximum activity intake of each unmonitored isotope. 

Table 7. Factors for unmonitored isotopes for the 

period 1953 through 1962 only. 

Isotope nCi/mg U in urine

Th-230 0.3

Th-228 0.002

Th-232 0.005

Ra-226 0.02

Pb-210 0.02

Ac-227 0.005

Pa-231 0.003

3.4 DETAILED TIMELINE AND HISTORY OF K-65 MATERIAL

Beginning March 22, 1949, the K-65 residues from MCW were sent to Lake Ontario Ordnance Works 

(LOOW) to be stored for future shipment (Heatherton 1949). The initial shipment was stored under 

canvas in the open; later shipments were placed in an igloo, but some drums remained stored beside 

the road (Heatherton 1950). Starting in late 1950 and continuing into 1952, much of the drummed 

material was transferred to a modified water tower that was 165 ft tall. This operation was performed 

by National Lead Company of Ohio (NLO) personnel from FMPC. In December 1950, it was reported 

that there was a backlog of 11,000 drums in the igloo and new drums coming from MCW were going 

directly into the tower (NLO 1977; LOOW 1951). The estimates of the amount of material in the tower 

included values of 3,874 t, 3791 t, or 1,757 t (TAC 1982, p. 31). The K-65 material in the tower was 

ultimately moved to underground concrete water tanks in Building 411 on the LOOW site during the 

decontamination and demolition project in 1984 and 1985 and listed as 3,891 t (Glenn 1987).

A number of drums were not transferred to the water tower; the number has been reported to be 

between 5,000 and 6,000 drums (Boback 1978; TAC 1982; Hershman 1952; Heatherton 1952b;

Vogel 1989). These drums were of varied sizes, from 30 to 55 gal. Between September 22 and 

November 14, 1952, 4,481 drums were transported from LOOW to FMPC (Schuman 1952; Strattman

1952; Boback 1978; Karl 1952). The lower number of drums actually transported is likely due to the 

consolidation of drums at LOOW due to drum conditions. 

In 1951, MCW began shipping K-65 material to FMPC rather than LOOW (Walden 1952, Vogel 1989). 

Between September 26, 1951 and April 30, 1952, 11,253 drums were sent from MCW to FMPC. A 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 32 of 76

total of 25,000 drums were shipped (Vogel 1989). As an example of the rate of these shipments, in 

late 1952, FMPC was receiving five railcars per week from LOOW and three railcars per week from 

MCW (Heatherton 1952a). Each rail car had the capacity to hold 96 drums. The shipments from 

MCW continued until January 1956 (Vogel 1989). This K-65 material was transferred to Silos 1 and 2.

From 1954 to 1957, K-65 material was transferred to Silo 2 from Q-11 processing at FMPC. The 

dates of the specific campaigns are as follows:

A campaign April 1954–June 7, 1954

Pilot Plant September 24, 1954–October 14, 1954

B Campaign October 23, 1955–February 3, 1956

C Campaign July 31, 1956–October 30, 1956

D Campaign March 7, 1957–April 2, 1957 (the date of the memo, Vogel 1989) 

Processing of the Q-11 ores would have preceded these dates by 1-2 weeks (Vogel 1989).

Miscellaneous other materials were placed in these silos (most likely Silo 2) as described in Table 8

(Vogel 1989).

Table 8. Materials placed in Silo 2 after Q-11/K-65 processing was complete. 

Amount Period Source Reference

5 drums 4/11/1959 Clean out residue Attachment 10 to 

Vogel (1989)

150 drums 6/10/59 Wastes and residues from K-65 activities Attachment 12 & 13 

to Vogel (1989)

Some ionium 6/20/60 Attachment 14 to 

Vogel (1989)

75 lbs 4/16/65 K-65 and Q-11 from Lucius Pitkin–New York City Attachment15 to 

Vogel (1989)

11 drums 8/15/66 7 drums from Middlesex Sampling Plant and 4 drums of 

cleanup material from hot raffinate building

Attachment 16 to 

Vogel (1989)

The total K-65 material in the two silos was 19,385,126 lb, including about 380,000 pounds from the 

processing of “rum jungle” ore from Australia in 1957 and 1958 (Vogel 1989, Attachment 18; Shaw 

1968; Robinson 1973). About 25% of the material was generated at FMPC (Robinson 1973). In a 

memo from W. E. Shaw to J. E. Carvitti and J. R. Lynch dated May 24, 1968 (Shaw 1968), an 

attachment dated February 18, 1959, identifies the source of the material in the two silos. The page is 

entitled “K-65 Residue in 1 & 2 Storage Tanks” with a note, “check on this date 2/18/59 from 101 

records and avg analyses.” The data in Table 9 is contained on the rest of the page.

Table 9. Inventory of K-65 material in Silos 1 and 2. 

Source Weight Drums Ra-226 (mg/t) Ra-226 activity (Ci)

NLO 5,363,712 12,130 300 804

TOA 2,463,363 6,460 750 924

MCW 11,558,051 25,260 500 2,890

This attachment is handwritten and signed by J. E. Carvitti and T. C. Feist. This document is one of 

the sources for the total weight of residues in the two silos. A memo that S. L. Hinnefeld wrote in

1982 discusses estimates of the amount of material and the 226Ra content of the two K-65 Silos 

(Hinnefeld 1982). The following values were discussed in this memo:

• 1,650 Ci – The memo indicates that this is the most commonly reported value (Hinnefeld 

1982). It is also quoted in “History of the Operation of the FMPC by NLO, Inc,” by Mead et al. 

(1985).
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 33 of 76

• 2,960 Ci – The memo derives this from a reported 226Ra content in a Draft SAR for the FMPC 

Waste Storage area (Hinnefeld 1982).

• 1,720 Ci – This number was derived from a 1957 memo that assumed the Belgian Congo ore 

was 40% U (Hinnefeld 1982). [A more common range was 50-80% (see Section 3.1); 60% U 

which would yield 2,580 Ci.] 

• 4,600 Ci – The memo indicated this value was found in a Technical Division memo and 

indicated it might be old and therefore suspect, but noted that there was no reference to 

support the value. The data that is discussed above from Shaw (1968) appear to be the 

source of this value. Several other pages in that document reference this value for the 226Ra

content of the K-65 silos. 

Vogel (1989, attachment 17, p. 21) has a value for the alpha activity for K-65 material of 1.19 × 104

dps/g. This document is titled "The Origin of K-65 Material" and was written by a Westinghouse 

employee (Vogel) in 1989. This value yields a total of 2,770 Ci in the 19,385,126 lb of K-65 material 

in Silos 1 and 2.

As can be seen from the information above and data in this timeline presentation, the exact activity of 

226Ra in the two silos is not known for certain, but the amount of activity is certainly in the range from 

1,650 to 4,600 Ci.

4.0

4.1 HISTORICAL BACKGROUND AND PROCESS LIMITS

RECYCLED URANIUM AT FMPC

The term Recycled uranium (RU) has been used in the nuclear weapons complex in a number of 

contexts; e.g., recycled scrap materials from uranium manufacturing operations, etc. as well as 

uranium that has been recovered from reactor fuel and targets that have been irradiated in production 

and/or research reactors. Throughout much of the history of nuclear weapons production there was 

no separate and distinct accountability category for RU, which contributed to confusion in developing 

the mass balance reports referenced in this report and discussed later. The term RU in this report is 

used only for reactor-irradiated and recovered uranium.

RU was recovered in the chemical processing plants at the Hanford, Savannah River, West Valley

and Idaho Chemical Processing Plant sites. RU was known to contain trace quantities of transuranic 

elements (referred to as TRU in the rest of this report) and trace quantities of fission product 

impurities. The primary contaminants were Pu-238, Pu-239, Np-237, and a fission product, Tc-99. 

The Pu-239, Np-237, and Tc-99 were the radionuclides of greatest concern and were tracked and 

documented for control purposes. The levels of contaminants were recorded in parts per billion (ppb) 

on a uranium mass basis. 

The first uranium feedstock introduced to the Fernald plant in 1953 was high grade ores (50-80% 

uranium), i.e. pitchblende ores from the Belgian Congo and other locations, which could be processed 

in a modified PUREX process directly to high purity uranium. Other uranium feed stock came from 

low grade ores, which was processed near the mines by mills that produced natural uranium in the 

form of yellow cake (U3O8) in the range of 60% uranium. The yellow cake required further processing 

by PUREX to a high purity state. When RU from the chemical processing plants was introduced to 

the process feed streams at Fernald it was a high purity oxide and was normally blended with natural 

uranium or with other existing feed stocks.

Relatively small quantities of RU (a recorded mass of 45 MTRU) from Hanford were received at 

FMPC as early as 1955 (2000a,b) but did not constitute a feed stream component of significance or at 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 34 of 76

all. The first significant quantity of RU was received by FMPC for processing on February 13, 1961, in 

a shipment of UF6 that was to be converted to UF4 (DOE 2000a) and introduced to the feed streams of 

the plant.

4.1.1

Table 10 details DOE’s estimate of quantity and characteristics of RU at FMPC. Approximately 218 g 

of Pu-239, 25.7 kg of Np-237, and 332 kg of Tc-99 were received at FMPC in 246,683 metric tons 

(MT) of RU during the history of the plant. (DOE 2000a). The historical average of contaminants in 

RU was approximately 1 ppb Pu-239, 104 ppb Np-237, and 1,346 ppb Tc-99. 

Major Contaminants in FMPC RU

A limit of 10 ppb total plutonium in RU was set by the nuclear weapons complex (primarily Oak Ridge) 

for use at Hanford (ORAUT 2010; and DOE 2000b). Initially no numeric limits were specified for Np237 or Tc-99. The plutonium limit of 10 ppb was set such that the radiological impurities did not 

exceed 0.1% of the alpha activity from natural uranium. Table 11 lists the activity percentages and 

illustrates that the activity ratios of the impurities vary as a function of the activity of uranium that 

changes with enrichment. Note that the percentage is higher for depleted uranium, but in each case 

Table 10. Average characteristics of RU processed at FMPC. 

Uranium 

enrichment

MTUa

receipts

Percent 

uranium 

receipts

Total 

Pu-239 

(g)

Pu-239 

(ppb)

Total 

Np-237 

(g)

Np-237 

(ppb)

Total 

Tc-99 

(g)

Tc-99

(ppb)

Enriched 60,181 24.4 207.9 3.5 19,048 316 328,740 5,382

Normal 89,649 36.3 4.1 <0.1 3,026 34 1,197 13

Depleted 96,853 39.3 5.7 <0.1 3,669 38 2,061 21

Total or average 246,683 217.7 0.9 25,742 104 331,998 1,346

a. MTU = metric tons of uranium.

the radioactivity from uranium alpha emissions far exceeded the radioactivity from plutonium alpha 

emissions.

Table 11. Activity in RU containing 10 ppb and 100 ppb 239Pu for various uranium enrichments. 

Uranium 

enrichment

Uranium 

activity 

concentration

(pCi/μg U)

Plutonium 

activity

concentration 

in 10-ppb RU

(pCi/μgU)

Plutonium 

activity

concentration 

in 100-ppb RU

(pCi/μgU)

Fraction of total 

alpha activity 

from Pu-239 in 

10-ppb RU (%)

Fraction of total 

alpha activity 

from Pu-239 in 

100-ppb RU (%)

Depleted 0.422 6.175E-04 6.175E-03 0.154 1.54

Natural 0.685 6.175E-04 6.175E-03 0.09 0.90

1% enriched 0.976 6.175E-04 6.175E-03 0.06 0.63

2% enriched 1.616 6.175E-04 6.175E-03 0.04 0.38

4.1.2

As will be explained later, from the beginning of operations urinalyses for uranium was conducted 

extensively for purposes of controlling biological heavy metal concerns. The presence of RU 

contaminants was recognized but was not specifically analyzed due to the anticipated low levels and 

related low hazard contribution. The approach to bound the maximum credible dose from these 

contaminants is to add a default concentration of each contaminant to the extensive uranium-in-urine 

data. For this reason, it is not necessary to precisely quantify the mass balance of RU and material 

flow of RU in the weapons complex to establish a bounding estimate of the contaminant concentration 

in RU at FMPC. It is necessary to have an understanding of the materials flow into the processes but

basically to know the relative concentration in each of the major process streams, such that bounding 

level concentrations can be established. Figure 5 illustrates the flow or RU into and out of FMPC.

Mass Balance, Material Flow, and Pu-239 Levels in RU within the Weapons Complex 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 35 of 76

Figure 5. RU mass flow in and out of FMPC. 

It is true that the DOE documents pertaining to mass flows of RU in the weapons complex include 

some discrepancies, but these discrepancies in mass flow do not preclude use of the data to define 

bounding assumptions. The inconsistencies in the amounts of RU in various reports can be attributed 

primarily to definition for protection purposes, i.e. many plants simply classified and treated as such all 

the uranium in the plant as RU once the RU was introduced in the processes - due to mixing, 

blending, etc. In addition, the availability of data at the time that each report was prepared, the scope 

F 

E 

R 

N 

A 

L 

D 

Y-12

130 MT

0.3 g Pu

40 g Np

800 g Tc K-25

20 MT

< 0.1 g 

Pu

<10 g Np

 

Y-12

9530 MT

0.2 g Pu

40 g Np

173 g Tc

PORTS

1720 MT

7.1 g Pu

300 g Np

90 g Tc

PORTS

850 MT

0.5 g Pu

60 g Np

300 g Tc

PGDP

36690 MT

24.9 g Pu

4.7 kg Np

63.1 kg Tc

K-25

2720 MT

2.8 g Pu

20 g Np

300 g Tc

PGDP

59700 MT

38.9 g Pu

6.1 kg Np

17.2 kg Tc

RMI

67320 MT

113.4 g Pu

10.2 kg Np

185.2 kg Tc

RMI

57800 MT

74.3 g Pu

6.8 kg Np

121.3 kg Tc

West Valley

620 MT

1.8 g Pu

70 g Np

1.5 kg Tc

Weldon Spg

45363 MT

2.3 g Pu

0.3 kg Np

6.9 kg Tc

Hanford

22770 MT

68 g Pu

7.7 kg Np

156.4 kg Tc

SRS

6390 MT

14.4 g Pu

1.4 kg Np

24.9 kg Tc

RFP

1320 MT

< 0.1 g Pu

<10 g Np

<100 g Tc

Idaho

2 MT

0.1g Pu

<10 g Np

<100 g 

Others

48140 MT

7.6 g Pu

2.8 kg Np

2.7 kg Tc

Weldon Spg

10970 MT

2.4 g Pu

330 g Np

7.2 kg Tc

Hanford

48380 MT

23.2 g Pu

6.7 kg Np

8.6 kg Tc

SRS

32150 MT

12 g Pu

1.9 kg Np

34.3 kg Tc

RFP

5350 MT

<0.1 g Pu

<10 g Np

<100 g Tc

Idaho

84.4 MT

<0.1 Pu

<10 g Np

<100 g 

Others

37880 MT

14.5 g Pu

3 kg Np

38.4 kg Tc

West Valley

< 0.1 MT

<0.1 g Pu

<10 g Np

<100 g Tc
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 36 of 76

of each report, and the degree to which conservative assumptions were used contributed to the 

apparent discrepancies in mass flow. However, the DOE (2003a) mass balance report documented 

the quantities of contaminants leaving the RU generating sites (primarily Hanford). These numbers, 

though of interest in terms of understanding the flow and quantities of contaminants were not used in 

determining the recommended defaults. The defaults were determined by analyses of concentrations 

in the various processes and reported in Appendices F and F.1 in the Ohio Mass Balance report 

(DOE 2000a). The defaults were chosen such that they bound the highest values of all the 

processes, with the exception of the concentrations in the GDP tails and wastes.

The concentration of the contaminants also varies slightly with the source of the data. For example, 

the mass concentration of plutonium in DOE (2000a) is 0.9 ppb; for the complex-wide rollup, the mass 

concentration is 3.7 ppb; in DOE (2003), the mass concentration from the primary sites is 4.2 ppb. 

These values are essentially the same and all well below the 10 ppb control guideline specified for 

Hanford.

Further discussion of this issue and the DOE studies pertaining to uranium mass flow to and from 

FMPC are described in section 4.4.

4.1.3

The chemical separation of fission products and activation products from RU in the uranium 

processing plants was excellent in order to meet the uranium purity requirements, but did allow ppb 

levels of contaminants. Routine process chemical analysis at Hanford included gross beta and 

gamma measurements. These results were compared to the baseline measurements from aged 

natural uranium (NU) (DOE 2000b). Limits on these other contaminants were set so that the beta 

activity in RU was less than 100% of the beta activity in comparison with aged NU and gamma activity 

in RU was less than 100% (average of a ten lot sample with no sample > 300%) of the gamma activity 

in aged NU. The numeric value for these limits was 0.685 pCi/μg NU which is equal to 310 μCi/lb NU. 

With the advent of gamma spectrometry in 1967, the limits included isotopic levels as listed in 

Table 12.

Other Contaminants in RU

Table 12. Limits for other contaminants in RU. 

Isotope(s) Limit (μCi/lb U) Limit (pCi/μg U)

Gross beta count <100% beta activity of NU <0.685

Gross gamma count <100%–300% gamma activity NU <0.685

Zr/Nb-95 15 0.033

Ru-103/Ru-106 50 0.11

All other isotopes 2 0.0044

Gamma spectrometry performed at the Hanford Site did not identify 137Cs in RU, although trace 

quantities were identified in the analyses of dust collector data taken at FMPC. Strontium-90 and 89

could have been among the other contaminants, but would be just 0.6% of the gross beta activity from 

NU and 1% of the activity from DU. Detectable levels of Zr/Nb-95 and the ruthenium isotopes (Ru/Rh106) have been reported in RU samples from the chemical process plants. It was reported that 

additional reprocessing was necessary, primarily due to ruthenium contamination, until the product 

met the specifications. (DOE 2000b).

Some analyses have indicated that thorium isotopes were present in trace quantities. Insufficient data 

exist to establish a general default value for all sites, but modeling using the ORIGEN2 code has 

indicated the possible levels. For the major primary source (the Hanford Site), the analysis has 

indicated a default value for thorium of 10 ppm U. The values in Table 12 represent the acceptance 

limits used by Hanford. Specific analytical information for trace concentrations of other fission 

products as documented in Hanford references (DOE 2000b; ORAUT 2006a) are listed in Table 13.. 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 37 of 76

Table 13. Other radioactive contaminants in RU at Hanford. 

Element/isotope Observed range

Plutonium (primarily Pu-239) <1–2 ppb U

Neptunium (primarily Np-237) 0.04–0.16 ppm U

Thorium 8–10 ppm U

Technetium 3–4 ppm U

Ru-103/106 <6 μCi/lb U

Zr/Nb-95 <4 μCi/lb U

Other gamma emitters excluding Tc-99 0.09–0.75 μCi/lb U

4.1.4

The gaseous diffusion process concentrated the RU contaminants (primarily due to lack of volatility of 

the fluoride compounds as compared to those of uranium) in tower ash and other relatively low 

volume residues in the UF6 conversion and handling processes. These ash and residues also 

contained significant amounts of uranium, which required that they be recovered according to 

economic discard criteria. In 1972 the decision was made by the AEC to process these materials 

from the Paducah Gaseous Diffusion Plant (PGDP) and the other GDP sites. Fernald was directed to 

process most of these materials. This decision introduced RU contaminants to Fernald with 

Plutonium-Out-of-Specification (POOS), in other words above the established specification of 10 ppb. 

POOS feed stock from PGDP was sent to FMPC for reprocessing with the prior knowledge that the 

material was above the established contamination limit. The 1970s and increased amounts in the 

1980s GDP shipments accounted for a major portion of the total TRU inventory at FMPC, essentially 

doubling the plant inventory over this time period (Bassett et al. 1989 and DOE 2000a). 

Tower Ash and Plutonium Out of Specification

Consequently, exposures to RU contaminants greater than the 10 ppbU could have increased 

significantly starting in 1973.

According to the Fernald Closure Project (FCP) – Internal Dosimetry TBD (Tomes 2001), most of the 

POOS uranium on the site in 2001 contained less than 80 ppb plutonium. However, an examination 

of the statistical data sheets in the F and C Appendices of DOE (2000a) indicates that maximum 

values in some subgroups (as determined by the 95th percentile of a log normal distribution) ranged 

up to approximately 400 ppb Pu. These values are summarized in Table 14. A notable exception is 

the tower ash from PGDP, which had a mean value of 1732 ppb Pu (range of 0.6 to 3,505). This 

maximum-level POOS material was handled as a recognized short-term risk during feed preparations 

and blending operations. Additional protective precautions were applied during these short term 

tasks. These precautions included airline respiratory protection, protective clothing over normal plant 

clothing which was comprised of head covering, gloves, shoe covers. Tape was applied at ankles 

and sleeves. Weekly urine bioassay was required of those involved. (Author unknown) All other 

process subgroups as a result of the POOS insertion had 95th percentile values less than 400 ppb 

Pu. Thus, the short duration of the injection of the subgroup 10A materials into plant inventory 

materials with the immediate drop to subgroup 10B levels of 35 ppb Pu (range 2.5 to 53.1 ppb Pu) 

make the default level of 400 ppb recommended in this report adequate to ensure claimant 

favorability. 

Table 14. RU contaminant levels by process subgroups.

a

Subgroup Description

Pu ppbU 

@95%

Np

ppbU@95%

Tc ppbU 

@95%

1A Miscellaneous 133 3,692 3,060

1B Miscellaneous – minor offsite 1.6 179 1

2 UF6 source UF4 (GDP tails) 1.2 84. 171

3 UF6 source metal and scrap 0.01 4.2 15

4 Normal U product, residues, and scrap 0.12 139 33

5 Enriched UF6 source products/res. 3 109 15,410
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 38 of 76

Subgroup Description

Pu ppbU 

@95%

Np

ppbU@95%

Tc ppbU 

@95%

6A UO3 PUREX source (A508)(unblended) 4.6 655 14,119

6B LEU products A508 UO3/UF4 (low cross) 4.1 643 19,732

6C LEU products A508 UO3/UF4 (high cross) 34 1,198 4,823

6D A500 coded enriched residues 7.6 246 2,701

6E SRS UNH 37.8 nmb nm

6F SRS UO3 – not shipped to FMPC 4.6 nm nm

7A A508 based derbies 17 513 2,870

7B A508 based ingots and metal 1.7 433 133

8 Enriched MgF2 342 4,417 4,619

9 Incinerator ash and scrap residues–GDPs 283 5,116 474

10A Tower ash and decontamination residues 1732 17,277 4,445

10B UO3 from tower ash 34.9 842 4,154

11 Waste residues 152 10,666 6,991

a. *Though the results in the table are all reported in ppb U, this measure may appear less meaningful in subgroups in 

which there is a reduced amount of uranium (in single digit percentage levels) such as subgroups 8 and 11, in which the 

MgF2 and chemical plant raffinates accumulated some isotopes, but were relatively low in uranium by design, but still 

continued to be a process stream to recover uranium above the economic discharge levels. Note that even with 

decreased U levels the contamination default can still be a ratio of U.

b. nm – not measured.

Over time, much of the POOS material was blended with the natural uranium feed stock and with RU 

of typical concentration. This was done to maintain the plant-wide limits that had been determined to 

provide an adequate level of safety, i.e., staying near the average complex-wide plutonium 

specification of 10 ppb). Workers handling the POOS materials directly prior to its being blended with 

other materials (Plant 1 and other locations) were provided with special protection. As a background 

comment, a small amount of this POOS material from the GDP was inadvertently stored for a period 

of years. The discovery of this material in 1985 was categorized as an unusual event with the 

material being subsequently processed and disposed. 

After processing the PGDP POOS, the blended uranium oxide resulted in temporary storage of 

uranium with a maximum concentration of Pu of 43 ppb U. In addition, several of the uranium 

purification processes resulted in concentrating the TRU contaminants in waste streams. Formal 

investigations were conducted that traced the TRU contaminants through the various plant processes; 

one such study in 1975 defined contaminant affinity for the metal production furnace mold materials 

(some TRU studies are referenced in DOE 2000a). These analyses demonstrated that the extraction 

and metal conversion processes concentrated plutonium, neptunium, and uranium daughter product 

contaminants in the raffinate waste of the liquid extraction columns and in the MgF2 from the thermite 

metal conversion furnaces. MgF2 slag from the thermite metal reduction process was also processed 

through a slag leach process and then through the chemical extraction plant. These concentration 

points or areas were identified by analytical methods in early years. The discussions in this report 

make use of studies of the data from those early reports (DOE 2000a). The recommended defaults 

account for all of the possible concentration points in the processes.

4.2 WORKPLACE AND PERSONNEL MONITORING ASSOCIATED WITH RU

The FMPC Health and Safety staff was aware that TRU and fission product contaminants were 

present in the recycled uranium. The biological concerns were thought to represent an increase of 

less than 10% in hazard level compared to uranium that had never been irradiated. Radiological 

controls were determined and adjusted based on routine air sample analysis and urine bioassay for 

uranium. 

Before 1986, routine TRU analyses was not performed for either airborne or urine activity, although a 

few examples of specific analyses of Pu on air samples were recorded and indicated the uranium to 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 39 of 76

be controlling. Exposure controls were set at levels that prevented chemical toxicity of uranium 

assuming that these controls were sufficient for all the radiological hazards (Bassett et al. 1989). 

Although the alpha activity from the TRU alpha emitters would have been collected and detected on 

the air samples, the reported results were all considered to be primarily uranium and compared to the 

Maximum Air Concentration (MAC) for uranium. Consequently bioassay measurements were not 

routinely performed on workers for the contaminants associated with RU until 1986. Note SRDB 4613 

referenced in Section 4.2.1 below. 

4.2.1

A review of the Oak Ridge Y-12 radiation protection program and internal dosimetry technology 

(Cofield 1959a; Scott 1964; Steckel and West 1966; West 1979) indicates that the internal dose 

technology, techniques, procedures, and philosophy used at FMPC were similar to those at Y-12. 

The fact that the Oak Ridge Y-12 Mobile In Vivo Radioactivity Measurements Laboratory (MIVRML) 

provided routine service to the FMPC in vivo internal dosimetry program is an indicator of the working 

relationship the FMPC staff had with the Y-12 program. However, prior to 1988 this technology did 

not provide detection for TRU at the very low levels that existed in the plant from RU contaminants. 

Typical results of the MIVRML were lung burdens for uranium and U-235, and data to determine 

thorium uptakes. Though TRU analyses were attempted, the limitations of the MIVRML, which 

included limited detection sensitivities for TRU isotopes and infrequent counts, restricted the ability to 

detect the anticipated levels at FMPC and in fact could not detect levels that met regulatory limits for 

chronic exposure to TRU isotopes. This lack of capability was a technological limitation of in vivo

counting within the nuclear industry in general during this period of time. Early in 1989 an in vivo

counting facility was constructed on the FMPC site to replace the mobile in vivo counting facility from 

Y-12. 

In Vivo Analysis Data

Before 1989 when DOE Order 5480.11 was issued, bioassay data at FMPC was not routinely used to 

estimate intakes and internal organ doses. Instead chest (lung) counting was used to estimate the 

fraction of a Maximum Permissible Lung Burden (MPLB) for uranium primarily and infer annual dose 

by multiplying by 15 rem/y per MPLB. Lung burdens for thorium were also determined and recorded. 

Workers were removed from the higher exposure work or placed on radiological work restrictions 

when in vivo results exceeded 75% MPLB. The in vivo counting frequency was set at once per year 

for the high exposure potential work groups (chemical operators), with the exception that workers 

exceeding 50% MPBB were counted more frequently. Annual exposure reports listed uranium lung 

burdens in percent of MPLB only and did not address the systemic radiological burdens. See SRDB 

4613, R. C. Heatherton paper, Occupational Health Experience With A Contractor Uranium Refinery, 

April 1975 for a discussion of the results of this program during the early time periods.

In summary, direct in vivo lung monitoring began in 1968 with the MIVRML and continued in 1989 with 

the FMPC IVEC counting facility until 2001. No RU contaminant analyses were reported in the in vivo

records before 1989. Although there would have been some ability to detect TRU materials in the 

range of 0.1 to 10’s of nCi depending on the radionuclide, the in vivo counts were not performed with 

a consistency or frequency to be of significant value in TRU dose reconstruction. 

4.2.2

Measurements of uranium in urine were compared to limits based upon preventing toxic effects from 

heavy metals. A more formal program of internal dosimetry was introduced in 1986 at the change of 

contractors for the FMPC site. An immediate emphasis was placed on a more formal evaluation of 

the transuranic materials and fission products. From April 1986 to 1989, 675 bioassay samples from 

441 workers were collected for plutonium analysis. “Those samples were collected primarily from 

workers who were expected to be and, in some cases, were actually involved in POOS processing 

and cleanup campaigns in Plants 4 and 8.” Only 10 individuals exhibited quantities of plutonium in the 

Urine Analysis Data
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 40 of 76

urine above 0.02 dpm/sample. (This contractual urine MDA was determined to be unrealistically low 

and later adjusted to 0.13 dpm/sample for Pu urine analyses). Further investigation (follow up 

samples) revealed that 3 of the 10 were below the specified detection limit of 0.02 dpm per sample. 

All of the 10 individuals were flown to Hanford, where in vivo counts were conducted for plutonium. 

All results were below the detection limit of the Hanford in vivo analyses. The conservative 

assumption was made by the site contractor that seven of the ten were technically at the urine MDA. 

Directions were given to “regard those results as valid indicators of exposure and to calculate the 

plutonium dose commitments to those workers based on these worst-case assumptions. The dose 

commitments are to be reported in each employee’s permanent exposure record.” (Bassett 1989). 

The FMPC radiation protection practices were adjusted to more rigorously account for those trace 

isotopes within the DOE radiation protection standards in effect at the time..

The information above is taken from the report of a DOE team of investigators and documents an 

organized effort by an incoming contractor to verify the previous assumptions that TRU exposures 

were not a major exposure source. The bioassay and in vivo results will be found in the claimant files 

and might be used by NIOSH dose reconstructors in some cases to assure bounding of doses. 

However, this data is insufficient to assign intakes/internal dose from contaminants in RU for all 

operational periods. The default intake analyses proposed in this paper are based upon recorded and 

maximized ratios of RU contaminants in uranium from which contaminant intake analyses can be 

performed.

4.2.3

Before February 1986, smears or air sampling filters were not specifically nor routinely analyzed for 

plutonium, neptunium, or thorium isotopes (Basset 1989), although these radionuclides would have 

been detected by the routine gross alpha counting. In 1989, several sets of air and surface smear 

samples from Plants 4 and 8 were analyzed for total uranium, 239Pu/240Pu, 238Pu, 237Np, 228Th, 230Th, 

and 232Th. As stated above, a few examples of specific analyses of Pu on air samples were recorded 

at earlier times and indicated the uranium to be controlling. Table 15 lists the results of the air 

samples in derived air concentrations (DAC) from these analyses. The DAC ratios are provided not 

for dose reconstruction purposes, but only to indicate the relative measured levels of TRU 

contaminants in the plant in the 1989 time period. However, the ratios lend credence to the default 

assumptions proposed in this report, which are intended to account for the unmeasured TRU in the 

plants.

Air Sampling Data

Table 15. Derived air concentrations from 1989 RU isotopic analysis.

Isotopes Derived air concentrations

Pu-238, Pu-239, Pu-240 0.1-1% of Type S

Np-237 <0.03% to 0.1% of Type S

Th-228, Th-232 Ranged up to 3% of the Type S

Th-230 6 to 10% of Type S

Note: These initial results indicated that for those samples where the activity fraction 

of each radionuclide could be compared (generally those with a total alpha activity 

exceeding 1 × 10-12 μCi/cm3), uranium was the controlling activity in air samples. In 

addition the results were reported in percentage of the Derived Air Concentration limits 

in effect at the time, which were also in the range of 1 × 10-12 μCi/cm3. 

During the next several years, a major sampling and analytical program examined hundreds of 

process samples, air samples from the process areas, and biological samples for plutonium, 

neptunium, thorium, technetium, and uranium. The air monitoring program was used to establish 

work controls, such as respiratory protection requirements for workers, and was not routinely used to 

establish internal intake or exposure estimates although DAC-hr exposure evaluations were placed in 

the personnel dosimetry files for some specific operations, i.e. thorium materials. 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 41 of 76

4.3 RU IN WASTE STREAMS

4.3.1

This section deals only with raffinates and waste products during the period (post 1961) associated 

with processing RU. (Raffinate waste products associated with the processing of high-grade 

pitchblende ores and processed yellowcake from the uranium mines and mills are covered in 

Section 2.0 above.) The primary defaults recommended in this report for trace-level contaminants 

that accompanied RU rely upon routine bioassay analyses for uranium. There are processes in which 

concentrating mechanisms existed, and in which the uranium in those process streams will be low by 

design. Table 16 lists the processes that occurred at FMPC (DOE 2000a). These are process 

streams, determined to be waste on the basis that the remaining uranium is “below the economic 

disposal limit” (EDL) – but not zero.

Process Raffinates and Other Wastes Below the Economic Disposal Limit (EDL)

Table 16. Processes that concentrated RU contaminants.

Process step/activity Explanation/comments

Chemical extraction 87.2% initial fed Pu and 41.6% initial fed Np reports to the UO3 product stream

UF6 to UF4 Concentration in heel of constituents and remains with container

Hydrofluorination Potential Tc vaporization

Reduction 46% Pu and 63% Np report to MgF2 slag

Vacuum casting 5% Pu and 64% Np report to crucible and mold residues

Metal pickling Surface residues (higher in Pu and Np) dissolve in pickling acid

Machining Top Crops greater than metal product in feed

The predominant process streams of concern on the basis of low uranium content are enriched MgF2

from the metal thermite process, a variety of raffinates, normal and depleted MgF2, sump cakes, etc. 

from liquid extraction processes. Enriched MgF2 was processed 1) for uranium recovery, 2) reuse, 

and 3) disposal as waste, while the raffinates were treated or processed and disposed in silos, drums 

and/or disposal lagoons. On the surface the default approach of adding RU contaminants to the 

uranium intake determined from uranium bioassays for workers involved with these processes can 

appear to be less reliable. The following discussion presents a description which explains the choice 

of the default ratios as claimant-favorable.

4.3.2

During the early period from 1953 to 1968 (including the RU period) air sampling time weighted 

exposure studies were performed throughout the plants. These studies illustrated the relationship 

between high concentrations and short exposure times to other more chronic exposures. The studies 

performed in the work locations defined as raffinate areas were generally lower in concentrations than 

those areas with higher concentrations of uranium by a factor of 10 or higher. DWE levels were 

generally in the 0.1 MAC levels or at the MDL. These data is shown in Section 3.2.

Daily Weighted Exposure (DWE) Studies

Even though the raffinates may have contained a relatively high ratio of RU contaminants to the 

uranium content compared to other plant locations, the air concentrations in the raffinate area were 

quite low. The lower concentrations in the Plant 3 Raffinate areas are the result of these processes 

being liquid based and totally contained within closed piping systems. These enclosed systems would 

have provided very little exposure to the workers, which the Daily Weighted Exposure studies 

demonstrate. The primary source of exposure occurred in the Plant 2 area where ore dumping 

operations were conducted. From the information above, exposure in the raffinate area was a factor 

of 20 less than in the Digestion areas, yet the MPC, based upon the measured isotopic composite, 

was only a factor of 15 smaller. For a specific example, if the exposure in the Digestion area were 

assumed to be at the U-Nat MPC of 220 dpm/100 m3, the corresponding exposure in the Raffinate 

area would be 11 dpm/100m3, less than the MPC calculated for the raffinate mixture. 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 42 of 76

Since the personnel of Plant 3 were rotated in the various jobs it was assumed that over a period of 

time each operator within a group would receive a similar exposure, i.e., no class of operators were 

assigned exclusively to the raffinate areas. This would tend to result in exposures to uranium that 

combined Pu contamination levels in the <1 ppbU with those in the range approaching 400 ppbU. If 

the Claimant were assumed to work in the Digestion areas full time, disregarding any work in the 

Raffinate area, the dose reconstruction recommendations listed in Section 4.3 below are sufficiently 

favorable to claimants to bound the potential exposures from the raffinate area based upon uranium 

intake analyses. 

The DWE Exposure Study reports from 1968 and 1969 (Ross, Leininger, and Lawrence 1968; Ross, 

Leininger, and Zimber 1969) also included data from the urine sampling program. The following 

paragraph taken from the 1968 report documents the NLO policy and basis for their reporting levels.

AEC Manual Chapter 0502 requires that all employees who are shown by bioassay to 

have one-half of one permissible body burden to be reported. National Lead Company 

of Ohio has arbitrarily chosen to report all employees whose urine results average 

0.050 mg uranium per liter over a six-month period. This is believed to represent a 

much smaller uranium burden than the one referred to in the AEC Manual.

Only rarely did any worker exceed the value representing one-half of the permissible body burden and 

most of the results were less than the value used for internal reporting, which was 0.025 mgU/L. 

These bioassay results are relatively low considering the results of the Exposure Studies, but reflect 

the conservative approach that was used in the study results to disregard the use of respirators, which 

in the higher exposure jobs was required. The uranium urinalyses results demonstrate that reportable 

uranium exposures occurred in all areas, including those identified as “raffinate” areas. Thus, RU 

contaminant intakes can be calculated by default ratios to positive uranium urine sampling results.

4.3.3

The following mitigating considerations also support the default recommendations:

Other Factors Associated with Recycled Uranium in Waste Streams

1. The primary source of RU from the reactors was generally not initially submitted to chemical 

extraction due to the high purity of this product. Thus, the levels of RU contaminants in the 

process areas will be higher prior to the concentration mechanisms.

2. Only a small portion (approximately 10%) of the process streams (scraps and process rejects, 

etc.) are processed by dissolution and chemical purification, and are diluted then with non-RU 

sources. Hence, the total quantities of RU contaminants in the processes that generate the 

raffinate streams are lower than in the other process areas.

3. The raffinate process has two aspects that minimize worker exposures: (1) being enclosed in 

process vessels and piping, which provide containment, and (2) the wet or moist material 

forms.

4. The raffinate process is a waste stream and is directly disposed . Generally, there is no 

continuing or further processing of the materials, which results in minimization of the exposure 

potential. 

4.4 RU MASS FLOWS AND DISCREPANCIES IN MASS FLOW REPORTS

In the late 1990s, DOE recognized the need to better understand the internal dose potential of 

contaminants from the historical mass flow of RU in the weapons complex. Reports were generated 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 43 of 76

by each of the nine groupings of DOE sites or areas that were involved with RU. DOE (2000a) covers 

the Ohio sites, including the FMPC, and it is the primary source cited in this report. 

Table 10 shows the average contaminant concentrations of the major contaminants in the RU that 

was received at FMPC during the period when RU was processed (1961-1999). 

Process subgroup evaluations are summarized in Table 14. In many instances, the data from the 

process subgroups did not conform well to normal or lognormal distribution patterns so “bootstrap” 

analysis was employed. The results of these statistical analyses are presented in the DOE 2000a 

report. This statistical technique uses a sampling approach to approximate representative values for 

the data set (DOE 2000a). For purposes of this report and default analyses, the lognormal distribution 

assumption and defaulting at the 95th percentile level was used to assure a more conservative default 

for all plant workers for the time period of the GDP wastes processing (1973-1986).

An additional study was published by the DOE Office of Security in May, 2003 (DOE 2003), with the 

specific purpose of resolving the recognized inconsistencies in the mass flows among the nine site 

reports published in 2000. Table 3-4 of DOE (2000a) contains the data related to receipts and 

shipments at FMPC taken, in part, from WMCO (1986). However, DOE (2000a) defines “recycled 

uranium” at Fernald as all uranium received or shipped after 1962, while the receipts and shipments 

listed in WMCO (1986) were total uranium. The DOE (2000a) report acknowledges that the mass for 

shipments of RU exceeded the mass for receipts of RU “due to limiting the envelope for data to after 

recycled uranium started, i.e., not including the years prior to 1962 at the FEMP.” Due to blending 

during chemical processing, any uranium received prior to 1962 would likely have become RU if it 

were shipped after 1962 and therefore the apparent mismatch in RU receipts and shipments is a 

logical outcome. 

The DOE Ohio Field Office report (2000a) discusses the difficulty of determining exact amounts of 

recycled uranium moving between the site and introduced the idea of “potentially recycled uranium for 

Fernald.” The report discusses the differences between the complex-wide rollup report and the site 

specific report, pointing out that although the amount of RU dropped from 246,000 MTU from the Site 

report to 55,419 MTU in the DOE complex-wide rollup, the amount of plutonium only dropped from 

217 g to 201 g. The observation was made that “the majority of the constituents are contained in a 

relatively small mass of uranium.” Examples of these mass balance inconsistencies are listed in 

Table 17. However, it should be remembered that the mass flow discrepancies do not influence the 

default assumptions based on TRU to uranium ratios, which were determined by specific analyses of 

process streams.

Table 17. Examples of uranium mass balance inconsistencies in FMPC receipts. 

Source

RU quantity

(MTU)

Pu-239

(g)

Np-237

(g)

Tc-99

(g) Comments

WMCO 1986 606,932 NR NR NR 1953–1986

DOE 2000a 246,683.1 217.7 25,742.1 331,998.1 1962–1999a values used in this report

DOE 2000a 55,419 201 18,378 328,981 1962–1999 complex-wide rollupb

DOE 2003 17,966 74.3 5,735 71,000 Revision by DOE Office of Securityc

NR = not reported. 

a. Includes receipts from both primary and secondary sites as well as an assumption of contamination of existing 

inventories.

b. This complex-wide rollup summary also includes primary and secondary shipments, but does not include contamination 

of existing inventories from uranium ores processed on site and at uranium mills.

c. Includes receipts from primary sites only. 

It is true that the DOE documents pertaining to RU in the weapons complex include discrepancies, but 

these do not preclude the use of process data to define bounding assumptions. The inconsistencies 

in the amounts of RU in various reports can be attributed to the availability of data at the time that 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 44 of 76

each report was prepared, the scope of each report, and the degree to which conservative 

assumptions were used. See Section 4.1.2 for further discussion. The assumption that, beginning in 

1961, all uranium at FMPC should be treated as RU is a major contributor to the RU inventory 

discrepancies. This assumption is favorable to the claimant.

Just as the mass inventories are different in the various reports, the concentration of the RU 

contaminants is somewhat, but less, different in the various reports. For example, the mass 

concentration of plutonium in DOE (2000a) is 0.9 ppb. But the complex-wide rollup DOE (2003) 

defines this value as 3.7 ppb. The mass concentration in a specific batch of uranium could vary from 

near-zero to values slightly in excess the Hanford guideline value of 10 ppb. However, these 

shipments from the primary sites in excess of the 10 ppb limit were rare and exception approved.

4.5 SUMMARY AND RECOMMENDED DEFAULT VALUES FOR RU AT FMPC

Significant quantities of RU appeared at FMPC beginning in 1961, but there were no workplace 

indicators or bioassay monitoring for the contaminants in RU prior to 1986. Consequently, dose 

reconstruction for exposure to RU at FMPC is based on uranium bioassay results and should include 

default factors to account for exposure to the RU contaminants. The chemical forms of the RU 

contaminants are not known. The dose reconstructor should assume the solubility for the RU 

contaminants that matches the claimant-favorable selection made for the uranium, as recommended 

in ORAUT-OTIB-0060, Internal Dose Reconstruction (ORAUT 2007a). Table 18a lists the 

recommended mass concentration additions and activity conversions for the period after 1961 and to 

1973. The actual contaminant levels were much lower than these values from 1961 to 1973, since 

their basis is process stream analyses in the time period of the highest activities following1973. 

However, these are the defaults used in the dose reconstruction program for several years and will 

continue to be used for simplicity and conservatism. The defaults for the 1973 time period forward 

were adjusted, using a lognormal distribution at the 95th percentile. 

Table 18a. Recommended RU default values beginning in 1961 

and to 1973.

RU contaminant

Mass concentration 

addition

Activity concentration

(Bq/g U)

Pu-239 100 ppb U 232.7

Np-237 3,500 ppb U 92.5

Tc-99 9,000 ppb U 5698

Th (232)a 10,000 ppb U 0.04

Ru-103/106 50 μCi/lb 4,075

Zr/Nb-95 15 μCi/lb 1,222

Others (Sr-90) 2 μCi/lb 163

a. Th-228 should be assumed to be at 70% equilibrium with Th-232.

Table 18b. Recommended RU default values beginning in 1973 

and to 1986.

RU contaminant

Mass concentration 

addition

Activity concentration

(Bq/g U)

Pu-239 400 ppb U 931

Np-237 11 ppm U 291

Tc-pp 20 ppm U 20,000

The indicated addition for each RU contaminant should be added for each gram of uranium in the 

urine.

The defaults for the time period of 1973 to 1986 should be those presented in Table 18b above, which 

represent the time period of processing the waste materials from the Gaseous Diffusion Plants (and 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 45 of 76

specifically the PGDP), which created a major increase in contaminant levels in all processes at 

Fernald. The other contaminant defaults should remain the same as in Table 18a. 

5.0

This report describes the average DWE hazard assessment process used at FMPC and explains how 

the DWE data can be interpreted and used to calculate inhalation and ingestion intake rates of 

thorium for use in dose reconstruction. 

USE OF DAILY WEIGHTED EXPOSURE REPORTS FOR ESTIMATION OF CHRONIC 

DAILY THORIUM INTAKE RATE

DWE data are especially important before 1968 when in vivo chest counting was introduced to the 

FMPC site. In addition, the concepts in this report can be used for dose reconstructions associated 

with raffinate or other waste-stream processing. 

Information on where and when various materials were processed is in ORAUT (2004). Thorium was 

handled in several FMPC plants before 1968 (Morris 2008). Thorium-bearing materials were 

received, sampled, and occasionally repackaged in Plant 1. Thorium fluoride was produced in Plant 4 

and, during the same period, reduced to metal and rolled into ingots in Plant 9. From 1959 through 

1963, thorium residues were processed in a sludge furnace in Plant 6. Occasional solvent extraction 

and metal casting occurred in the Pilot Plant beginning in 1964 and continuing into the 1970s. 

Table 19 illustrates where and when thorium is known to have been processed before 1968 (Morris 

2008). Plants that are not included in Table 19 are not known to have been involved in thorium 

processing before 1968.

Table 19. Where and when thorium was processed at FMPC before 1968. 

Plant 54 55 56 57 58 59 60 61 62 63 64 65 66 67

1 X X X X X X X X X X X X X X 

4 X X X N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A

6 N/A N/A N/A N/A N/A X X X X X N/A N/A N/A N/A

8 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A X N/A

9 X X X N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A

Pilot X X X N/A N/A N/A N/A N/A N/A N/A X X X X 

X = thorium was processed; N/A = thorium is not known to have been processed in that facility at that time. 

5.1 DAILY WEIGHTED EXPOSURE REPORTS

An average DWE is an assessment of the exposure potential of an individual with a specific job title or 

job description at a specific facility. The concept of the average DWE was introduced to the AEC 

complex by the HASL staff (Adams and Strom 2008). The earliest DWE report at FMPC was 

prepared by the HASL staff (Breslin, Loysen, and Glauberman 1953). The monitoring concept was 

transferred to the FMPC staff, proceduralized (Wing 1960), and used at FMPC since the beginning of 

site operations. The procedure required that an annual analysis be prepared for all production plants 

including the Pilot Plant, Development Machine Shop, Technical Laboratory, and Laundry. 

Nonproduction areas were also to be evaluated based on air dust samples from the Boiler and Water 

Purification Plant, Chemical Warehouse and Factory Stores, Decontamination Building, Mechanical 

Areas (Mechanical Building, Garage and Firehouse, and Heavy Equipment Building), and 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 46 of 76

Nonproduction Areas (Service Building, Administration Building, Personnel-Security Building, Health 

and Safety – Production Building, and Guard House).

In modern practice, the average DWE concept is similar to the time-weighted average (TWA) 

exposure that is in current use (ACGIH 1996), except the TWA assumes an 8-hour workday while the 

average DWE uses the actual workday duration. The peer-reviewed paper by Methner, Feng, and 

Utterback (2001) provides a description of the DWE method and data from FMPC.

DWE reports are summaries of data prepared by the FMPC Industrial Hygiene and Safety staff for 

use by management. The data in each report are in units of multiples of the MAC, which at FMPC 

was 70 alpha dpm/m3 of air until it was changed to 100 alpha dpm/m3 of air in 1963. The analytical 

technique applied to air dust samples was gross alpha counting. No isotopic information is available 

with gross alpha counting, and there are relatively few isotope-specific analyses in the record; 

therefore, the assumption about the radionuclide being measured was based on the material in 

process at that time. The reports are remarkably similar from year to year. For example, each report 

is typically between 30 and 70 pages long and includes the following information:

• Introduction containing a brief summary of the processes at the facility

• Description of the sampling and analysis method

• Summary of Data – Table I (see Figure 6 for an example), containing the average DWE for 

each job description in the facility, the number of workers employed in that job description, and 

an average DWE for the entire facility

• Summary of Data – Table II (see Figure 7 for an example), containing the average of the air 

dust sample concentrations for a specific operation or area

• Discussion, containing a detailed description of processes and controls

• Recommendations, containing general and specific recommendations, often in a form 

amenable to tracking

• Conclusions

• Appendix, containing Job Evaluation Reports.

DWE reports typically include a statement to the effect that respirators (filtering or air-supplied) were 

worn by some of the operations, but this was not taken into consideration in calculating DWE data, 

and it can be assumed that exposures of persons making proper use of prescribed respiratory 

protection were materially reduced from the calculated values. Another observation in some of the 

reports is that workers were rotated through jobs with the highest exposure potentials. Other data 

sources including interviews with former employees (ORAUT 2007b) confirm that use of respiratory 

protection was typical during tasks with high exposure. 

DWE reports are available for some of the FMPC plants for some of the years 1953 to 1971. A useful 

summary of the FMPC DWE data is available in the SRDB (Author unknown ca. 1968). The available 

data from the DWE reports for plants and years during which thorium was processed have been 

transcribed to spreadsheets.

1

 

1 The spreadsheets containing these transcribed data are available in the ”AB Document Review” folder. 

In the Pilot Plant during 1967 and 1969, no time-weighting information 

was recorded, possibly because the processes being characterized were in development and had not 

yet become routine. The transcribed DWE data represents six plants (1, 2/3, 4, 6, 8, 9), 14 different 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 47 of 76

Figure 6. Excerpt from a 1955 Plant 9 report (Stefanec 1955b) showing 

DWE for 119 employees by job description. The average DWE is in units of 

multiples of the MAC where the MAC is 70 alpha dpm/m3 of air.

years, 142 job descriptions, 361 operations, and 16,748 air samples. Air sampling duration ranged 

from 1 minute to 455 minutes, with the average being 63 minutes.

Figure 6 is an excerpt from a 1955 DWE report about FMPC Plant 9 (Stefanec 1955b) during a period 

when thorium processing was occurring. Figure 7 is an example of a Job Exposure Evaluation card 

for a Chemical Process Area Wet Area Helper from the same report. The card includes the time of 

the evaluated work shift. In almost all cases, the time on a card is 510 minutes or 8.5 hours. The 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 48 of 76

Figure 7. Excerpt from the appendix of a 1955 Plant 9 report (Stefanec 1955b) 

showing operations associated with a job description and the data used to assess 

each operation. Collectively this comprises a DWE for the job description.

Plant 9 Job Exposure Evaluations during 1955 were based on data from approximately 490 different 

air dust samples. 

Figure 6 also provides insight into the impact of tasks involving very high airborne dust concentrations 

on worker exposure. Air dust samples reflecting very high concentrations are often observed and the 

suspicion is raised that they reflect an out-of-control situation that precludes dose reconstruction. On 

the contrary, Figure 7 shows that as early as 1955 high air sample data were routinely assessed and 

preserved in the record. For example, one 15-minute task called "Dumping recycle oxide into predryer" was sampled on eight different occasions using a breathing zone air sampler. The highest 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 49 of 76

observed value was 125,061 dpm/m3 and the average was 54,960 dpm/m3. This task accounts for

49% of the exposure for this job description. Despite the task’s short duration and extremely high 

airborne dust concentration, its contribution to dose was assessed. A site-wide Dust and Fume 

Control Committee kept track of, routinely reviewed, and made recommendations about operations 

with the highest exposure potential (Starkey 1960–1962).The DWE reports also include an arithmetic 

mean value for the entire facility. For FMPC Plant 9 in 1955 the "average weighted exposure for the 

plant" is reported as 67.34 MAC.

5.2 CORRELATION OF DATA AND URINARY URANIUM EXCRETION RATE

At the 1958 Symposium on Occupational Health Experience and Practices in the Uranium Industry, 

Heatherton and Huesing (1958) presented a paper describing the correlation of weighted uranium air 

dust concentration with the urinary excretion rate of uranium at FMPC. This paper includes 

information beginning in 1952 and summarized below that provides a validation of the usefulness of 

DWE data as a predictor of excretion rates and, with the aid of biokinetic modeling, the intake rate. 

The following excerpt from the paper establishes the operational approach to exposure controls at the 

beginning of plant operation: 

We have always used urine sample results as an indication of the air contamination 

problem within the plant or on a job. A high frequency of urine sample results above 

0.050 mg/L within a particular job or plant has always required industrial hygiene 

investigation of the work to determine the cause. Field investigations have often led us 

to a source of exposure of which we were not aware at the time.

Thus, frequent urine sampling was combined with comprehensive air monitoring to provide the means 

of detection and control of uranium intakes. 

Figure 8, taken from the Heatherton and Huesing (1958) paper, illustrates the correlation of uraniumin-urine concentration with DWE, which was an analytical tool devised at the start of the programs for 

increased understanding of the effectiveness of the control mechanics. The Acme-Gridley operator 

job description is highlighted because of the abundant data available for this job – 120 urine samples 

were averaged to obtain the data point. Additional data associated with weighted air concentration 

levels ranging from 25 to 75 μg/m3 were used to develop a linear model. Data was constrained to this 

range because fewer air samples were collected for jobs and locations with lower air concentrations 

and intakes for jobs with highly elevated concentrations were likely to be biased by the mandatory use 

of respirators. The resulting model predicts an intake rate which would yield a urine concentration of 

0.030 mg U/L of urine when the worker is exposed to a DWE of 70 dpm/m3 (1 MAC). 

Beginning in 1958, urine sample reports were computerized and routinely sorted, listing the results by 

job number. This enabled comparison of urine samples with the DWE data for specific jobs. A urine 

sample concentration in excess of 0.050 mg/L was seldom found unless the DWE exceeded 1 MAC. 

This was based on an analysis of urine samples representing workers with negligible exposure who 

occupied administrative areas, workers who occupied production areas where the DWE was less than 

1 MAC, and workers who occupied production areas where the DWE was greater than 1 MAC. As 

expected, workers occupying areas where the DWE was greater than 1 MAC had the highest 

proportion of elevated uranium-in-urine concentrations. Conversely, workers in administrative areas 

where DWE values approach zero had very low uranium-in-urine concentrations. Workers occupying 

production areas where the DWE was elevated, but less than 1 MAC had urine concentration values 

tending between the two extremes. Table 20 presents the results.
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 50 of 76

Figure 8. Uranium-in-urine excretion as a function of DWE

(Heatherton and Huesing 1958). 

Table 20. Comparison of uranium-in-urine concentration with nominal DWE. 

Area

Number of 

urine samples

Nominal 

DWE

Urine concentration, mg/L (%)

<0.025 0.026 to 0.050 0.051 to 0.10 >0.10

Admin 267 << 1 MAC 94 4.9 0.7 0.0

Production 762 < 1 MAC 84 14 1.6 0.0

Production 318 >1 MAC 31 51 16 2.5

From this information the authors deduced a rule of thumb: If a worker's uranium-in-urine 

concentration is usually less than 0.025 mg/L and never in excess of 0.050 mg/L, the DWE does not 

exceed 1 MAC. 

The effect of respiratory protection on uranium-in-urine concentration was also reported. A specially 

planned job of "desludging" a salt bath furnace resulted in a DWE of 235 MAC on day 1, 5 MAC on 

day 2, and 2 MAC on day 3. Urine results for the 50 men wearing air-fed hoods were averaged. 

Before the job, the average uranium-in-urine concentration was 0.018 mg/L. At the end of day 3 the 

average concentration had increased only marginally to 0.038 mg/L. Fourteen of the men 

participating in this job wore dust-type respirators. For this group, the DWE during day 1 was 

28 MAC. Days 2 and 3 were the same as for the air-fed hood group. The workers wearing dust-type 

respirators began the job with an average concentration of 0.018 mg/L. At the end of day 3 the 

average concentration was 0.035 mg/L. Collectively, these data show the respiratory protection 

program provided adequate protection while performing a job with high exposure potential. 

This early study by the FMPC Health and Safety program (Heatherton and Huesing 1958) 

demonstrates the usefulness and reliability of using DWE data to bound intakes of contaminants in 

the workplace, when used as an integral element in the intake control program.
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 51 of 76

5.3 UNCERTAINTY

In a peer-reviewed publication, Davis and Strom (2008) evaluated DWE data and the supporting air 

sample data collected by the HASL staff at five Atomic Weapon Employer (AWE) sites: 

Electrometallurgical Co.; American Machine and Foundry Co.; Middlesex Sampling Plant; Horizons, 

Inc.; and Lake Ontario Ordinance Works. The data represented uranium, radium, and thorium 

operations performed from 1948 through 1954. The data from Horizons, Inc. was specific to thorium 

operations. The sampling methods and data analyses used by the HASL staff were the same as 

those used at FMPC. The engineered controls in these AWE facilities were similar to, but likely 

inferior to, those employed at FMPC, because FMPC was purpose-built in light of experience at 

earlier operating facilities. 

Davis and Strom (2008) determined that the DWE values were usable as inputs in dose 

reconstructions conducted in accordance with EEOICPA, but that the DWE values were limited by 

their lack of numerical uncertainties. The uncertainty associated with DWE values is a function of 

blunders (in other words, data entry or math errors), variability, and measurement uncertainty. 

Measurement uncertainty was judged to be trivial in comparison with the other factors. Variability in 

air sampling data can be the result of many factors, including representativeness of the sample, 

ventilation changes, sampler placement, process variability, and aerosol particle size. Using a Monte 

Carlo analysis, Davis and Strom retrospectively assessed the uncertainty and variability in the DWE 

values for 63 job titles, thereby eliminating the limitation of usefulness of the DWE data. 

Based on their analysis, Davis and Strom (2008) state it is reasonable to conclude that the upper 

95th-percentile value of the geometric standard deviation (GSD) distribution is about 4 and the upper 

99th-percentile value is between 7 and 8. This lends support to their generalization of assuming a 

GSD of 5 when a DWE concentration value is available but there is no information on uncertainty. 

For instances in which the TWA of the air sample data are not available, Davis and Strom (2008) 

determined that the upper 95th percentile of the site-wide air concentration data will almost always be 

favorable to the claimant when making compensation decisions. 

5.4 SELECTING A DAILY WEIGHTED EXPOSURE OR AIR SAMPLE CONCENTRATION 

VALUE FOR INTAKE RATE CALCULATION

Due to variations in job assignments, it is unlikely that unambiguous exposure scenarios can be 

defined for most workers. In addition, fugitive dust emissions have not been quantified except as the 

general air sample indicator. To ensure thorium intake potential is not underestimated, the highest 

DWE value associated with a specific job title or job description in the FMPC plant where thorium was 

handled for a specific year should be assigned to every worker in that facility. A GSD of 5 should be 

assumed.

If TWA data are not available, the upper 95th percentile of the air sampling data for that facility can be 

used instead.

If neither DWE data nor air sampling data are available for a specific year, or if they are judged to be 

inadequate or incomplete, DWE from the same facility for an adjacent period can be assigned. In 

these cases, a high value from an adjacent year should be selected. A GSD of 5 should be applied.

Table 21 summarizes the recommended DWE values for years and facilities in which thorium work is 

known to have occurred. These values represent the job description in each facility with the highest 

DWE value for that year.
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 52 of 76

Table 21. DWE values, in multiples of the MAC, recommended for use in estimation of thorium intake 

rates. 

Plant 1954 1955 1956 1957 1958 1959 1960 1961 1962 1963 1964 1965 1966 1967

1 (a) 16.4 4.3 1.5 1.0 2.1 0.5 0.6 1.1 0.6 0.7 0.6 1.0 1.2

4 4.5 3.2 0.9 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A

6 N/A N/A N/A N/A N/A 0.8 4.3 1.6 2.8 17 N/A N/A N/A N/A

8 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A 7.1 N/A

9 (b) 686 (b) N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A

Pilot Plant 5.9 (c) 2 N/A N/A N/A N/A N/A N/A N/A (d) (d) (d) 77

N/A = not applicable.

a. Data are not available. Use the value for Plant 1, 1955.

b. Data are not available. Use the value for Plant 9, 1955.

c. Data are not available. Use the value for Pilot Plant, 1954.

d. DWE data are not available. The value for Pilot Plant, 1967, is based on the 95th-percentile value of 18 air samples 

taken during thorium operations. Use this value for Pilot Plant, 1964 through 1966.

5.5 CALCULATION OF CHRONIC DAILY INTAKE RATES

The approach for intake rate calculations is based on the method defined in Battelle (2006). 

Two modes of intake are considered for routine dose reconstruction: inhalation and ingestion. The 

methods presented here require an appropriate DWE parameter to be selected by the dose 

reconstructor. That value might be the average DWE for a specific job description. As an alternative, 

the DWE value might be selected from a lognormal distribution fitted to average DWE data for the 

facility or site. 

For facilities in which thorium intake might have occurred, all activity should be assumed to be thorium 

in equilibrium between 232Th and 228Th. For example, if the thorium intake rate is 2 Bq/d, the 232Th

intake rate would be 1 Bq/d and the 228Th intake rate would be 1 Bq/d. 

5.5.1

The chronic daily inhalation intake rate Chronic INH for a worker can be calculated based on the 

appropriate DWE parameter, an assumed breathing rate BR and the hours worked in the job 

description per day Td as shown in Equation 15. 

Calculation of Inhalation Rate

) d

d(

7

5 ) dpm

Bq ( 60

1 ) d

hr () h () m m

dpm ()() d

Bq ( 3

3 )(

)*(

× × ×× × ∑

∑ = Op d T

CT

Chronic INH DWE MAC BR T d

d (15)

where 

DWE = DWE in multiples of the operational MACOp. A GSD of 5 is carried along with 

this value.

MACOp = MAC in dpm/m3, either 70 alpha dpm/m3 of air before 1963 or 100 alpha 

dpm/m3 of air beginning in 1963. 

Td = hours worked per day, typically 8.5 at FMPC

BR = assumed breathing rate per hour

1/60 = conversion to Bq from dpm.

This daily intake rate would be usable for dose reconstruction as a chronic inhalation intake rate if the 

worker were exposed to this concentration every day of the calendar year, but no employee was 

exposed every day of the year. To correct for the time away from the job due to weekends, the daily 

intake rate is multiplied by a factor of 5/7. This correction normalizes the calculated value and results 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 53 of 76

in an average chronic daily inhalation intake rate which is usable as input into the Integrated Modules 

for Bioassay Analysis (IMBA) program.

Assuming a typical breathing rate used in dose reconstruction of 1.2 m3

/hr and a typical 8.5-hour 

workday, Equation 15 simplifies to Equation 16. 

)( ××= )121.0( DWE MACOp d

Bq ChronicINH (16)

5.5.2

The chronic daily ingestion rate (Chronic ING) for a worker can be calculated based on the 

appropriate DWE parameter, the hours worked in the job description per day (Td), and factors defined 

by NIOSH. NIOSH (2004) describes three modes of ingestion intake. Mode 1 is associated with 

swallowed particles that were initially inhaled and deposited in the respiratory tract. This mode is 

incorporated in the International Commission on Radiological Protections (ICRP) biokinetic models 

and is not addressed in this paper. Mode 2 is associated with consumption of radioactive material 

that might fall out in a drinking cup where airborne radioactive material is present. It assumes the cup 

is receiving fallout 8 hr/d, but that value can be adjusted to reflect a workday of any length [1]. Mode 

3 is associated with pica, in which transfer of contamination from hand to mouth occurs. In addition, 

Mode 3 is based on the airborne radioactive material concentration, but it is unaffected by the length 

of the workday because airborne dust is assumed to be settling on surfaces 24 hr/d. Because Modes 

2 and 3 are based on airborne radioactive material concentration, they can be combined. 

Calculation of Daily Ingestion Intake Rate

Mode 2 assumes that the airborne radioactivity settles in a cup of a specified diameter with an 

assumed settling velocity over the course of an 8-hour day. In that case, the ingested activity is 

0.0985 m3

/d times the activity concentration per cubic meter. By dividing by 8, the workday length can 

become a variable, Td, in hours per workday. This makes the equation versatile to accommodate the 

typical 8.5-hour workday at FMPC. One other change is to adjust for time away from the job due to 

weekends, creating a chronic ingestion intake rate suitable as input to IMBA. This is done by 

including a factor of 5/7 in the equation. The resulting formulation for chronic daily ingestion rate from 

Mode 2 sources is Equation 17. 

) d

hr (8

) d

d(

7

5 ) d (0985.0) m

dpm

Bq ( 60

1 ) d

hr ()

m

dpm ()(

)(

3

3 )(

)*(

× ×× × × ∑

∑

= Op d T

CT

DWE MAC T

d

Bq Mode2 d

d

(17)

which simplifies to Equation 18. 

)( )1047.1( −4 DWE Op TMAC d ××××= d

Bq Mode2 (18)

where 

DWE = DWE in multiples of the operational MACOp. 

MACOp = MAC in dpm/m3, either 70 alpha dpm/m3 of air before 1963 or 100 alpha 

dpm/m3 of air beginning in 1963.

Td = hours worked per day, typically 8.5 at FMPC

The Mode 3 daily ingestion intake rate resulting from 1 workday is defined in NIOSH (2004) as the 

airborne activity concentration in activity units per cubic meter times 0.1 m3/d. Just as for Mode 2, this 

is adjusted to accommodate a 5-day workweek as shown in Equation 19 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 54 of 76

) d (1.0) m

d

d(

7

5 ) dpm

Bq ( 60

1 )

m

dpm ()() d

Bq ( 3

3 )(

)*(

× × ×× ∑

∑ = T Op

CT

Mode3 DWE MAC d

d (19)

.1( 1019 ) −3 DWE MACOp ×××=

Equations 18 and 19 can be combined to represent the chronic ingestion intake rate as shown in 

Equation 20. 

[( .1( 1047 ) .1() 1019 )] −4 −3 Chronic ING DWE MACOp Td ×+××××= (20) 

Assuming an 8.5-hour workday, which is typical at FMPC, this simplifies to Equation 21. 

)( )1044.2( −3 DWE MACOp ×××= d

Bq Chronic ING (21) 

where 

DWE = DWE in multiples of the operational MACOp. 

MACOp = MAC in dpm/m3, either 70 alpha dpm/m3 of air before 1963 or 100 alpha 

dpm/m3 of air beginning in 1963.

6.0

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional references, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database (SRDB).

ATTRIBUTIONS AND ANNOTATIONS

[1] Morris, Robert L. MH Chew & Associates, Inc. Principal Health Physicist. February 2008. 

This equation implicitly assumes the drinking cup is collocated with the worker at all times. It 

is reasonable to expect that a drinking cup would seldom be in areas where highest airborne 

concentrations occurred and would more likely be in a break area where exposures were 

lower. This assumption is, therefore, favorable to the claimant.
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 55 of 76

REFERENCES

ABRWH (Advisory Board on Radiation and Worker Health), 2010, Advisory Board on Radiation and 

Worker Health, Work Group on Fernald Site Profile and SEC, Friday, January 29, 2010, 

Hebron, Kentucky, January 29. [SRDB Ref ID: 93105]

ACGIH (American Council of Governmental Industrial Hygienists), 1996, Threshold Limit Values for 

Chemical Substances and Physical Agents and Biological Exposure Indices, American Council 

of Governmental Industrial Hygienists, Cincinnati, Ohio.

Adams, A. J., and Strom, D. J., 2008, "Uncertainty and Variability in Historical Time-Weighted 

Average Exposure Data," Health Physics, volume 94, number 2, pp. 145–160. [SRDB Ref ID: 

41616]

AEC (U.S. Atomic Energy Commission), 1955, Feed Materials Processing Center Plant 2 

Occupational Exposures to Airborne Contaminants, HASL-FMPC-13, New York Operations 

Office, Health and Safety Laboratory, Industrial Hygiene Branch, New York, New York, July 

21. [SRDB Ref ID: 3199]

Author unknown, 1968, Summary of Air Dust Surveys, 1952-1968. [SRDB Ref ID: 1529] 

Author unknown, Processing uranium Containing Plutonium Above target Limit, Safety Requirement

of Processing out of Specification Materials. [SRDB Ref ID: 33874]

Bassett A. C., et al., 1989, Health and Safety Provisions for Recycle Uranium Processing at the 

FMPC, Fernald Environmental Management Project, Fernald, Ohio, March 7. [SRDB Ref ID: 

33900]

Battelle (Battelle Team Dose Reconstruction Project for NIOSH), 2006, Site Profiles for Atomic 

Weapons Employers that Refined Uranium and Thorium, Batelle-TBD-6001, Richland, 

Washington, December 13.

Battelle (Battelle Team Dose Reconstruction Project for NIOSH), 2007, Default Assumptions and 

Methods for Atomic Weapons Employer Dose Reconstructions, Batelle-TIB-5000, Richland, 

Washington, April 2.

Boback, M. W., 1978, “Gamma Doses from K-65 Operations,” letter to R. C. Heatherton, National 

Lead Company of Ohio, Cincinnati, Ohio, June 21. [SRDB Ref ID: 3051]

Borak, T. B., 1985, “Calculation of Radon Emission, Dispersion and Dosimetry from K65 Storage 

Tanks at the Feed Materials Production Center,” Appendix A of M. W. Boback et al. 1987, 

History of FMPC Radionuclide Discharges, FMPC-2082 (revision to FMPC-2058); 

Westinghouse Materials Company of Ohio; Cincinnati, Ohio, May 1987. [SRDB Ref ID: 4270]

Breslin, A. J., Loysen, P. and Glauberman, H., 1953, Feed Materials Processing Center Pilot Plant 

Occupational Exposure to Uranium, U.S. Atomic Energy Commission, New York Operations 

Office, March 11. [SRDB Ref ID: 3192, p. 2]

Cofield, R. E., 1959, In Vivo Gamma Counting Method of Determining Uranium Lung Burden in 

Humans, Report No. Y-1250, Union Carbide Nuclear Company. May 22. [SRDB Ref ID: 686] 

Davis, A. J. and Strom, D. J., 2008, "Uncertainty and variability in historical time-weighted average 

exposure data," Health Physics, volume 94, number 2, pp. 145–160. 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 56 of 76

DOE (U.S. Department of Energy), 2000a, Review of Generation and Flow of Recycled Uranium at 

Hanford, DOE/RL-2000-43, Richland Operations Office, Richland, Washington, June 30. 

[SRDB Ref ID: 16494] 

DOE (U.S. Department of Energy), 2000b, Recycled Uranium Project Report, DOE 00-0001, Ohio 

Field Office, Fernald, Ohio, May 15. [SRDB Ref ID: 3644] 

DOE (U.S. Department of Energy), 2003, Recycled Uranium-United States Production, Enrichment, 

and Utilization, DOE/SOO-0003, May 2003. [SRDB Ref ID: 23595]

Glenn, R. D., 1987, “Bechtel Job No. 14501, FUSRAP Project DOE Contract No. DE-AC05-

810R20722, Radiation Measurements made during K-65 Residence Transfer, Code: 

7430/WBS: 202,” letter to J. T. Grumski (Westinghouse Electric Company), Bechtel National, 

Oak Ridge, Tennessee, February 24. [SRDB Ref ID: 17731]

Gray, D. E., editor, 1963, American Institute of Physics Handbook, Second Edition, McGraw-Hill, New 

York, New York.

Halcomb, R. N., and J. A. Huesing, 1957, Exposure Study of Plant 3 Personnel to Airborne 

Radioactive Dust, National Lead Company of Ohio, Health and Safety Division, Cincinnati, 

Ohio, December 31. [SRDB Ref ID: 41029]

Heatherton, R. C., 1949, “Unloading K-65 from Railroad Car at Modeltown,” letter to B. S. Wolff (U.S. 

Atomic Energy Commission), Linde Air Products Company, Kenmore, New York, March 25. 

[SRDB Ref ID: 3943]

Heatherton, R. C., 1950, Lake Ontario Ordnance Works Air Radon and Dust Survey, August - 

September 1950, LOOW-1, U.S. Atomic Energy Corporation, New York Operations Office, 

New York, New York, November 6. [SRDB Ref ID: 9078, p. 2]

Heatherton, R. C., 1952a, untitled letter to J. A. Quigley, National Lead Company of Ohio, Cincinnati, 

Ohio, September 5. [SRDB Ref ID: 2997]

Heatherton, R. C., 1952b, “Visit to Lake Ontario Storage Area,” letter to J. A. Quigley, National Lead 

Company of Ohio, Cincinnati, Ohio, September 3. [SRDB Ref ID: 2996]

Heatherton R. C., and J. A. Huesing, 1958, “A Uranium Refinery and Metal Plant Urine Program and 

Data,” Symposium on Occupational Health Experience and Practices in the Uranium Industry, 

Held in New York City, October 15 – 17, 1958, HASL-58, U.S. Atomic Energy Commission, 

New York Operations Office, New York, New York, September. [SRDB Ref ID: 7886, p. 81]

Hershman, H. J., 1952, “Transfer of Drummed K-65 to Fernald,” memorandum to C. L. Karl, U.S. 

Atomic Energy Commission, Tonawanda Sub Office, Tonawanda, New York, July 22. [SRDB 

Ref ID: 16408]

Hinnefeld, S. L., 1982, “Radium-226 in K-65 Tanks,” memorandum to M. W. Boback, National Lead 

Company of Ohio, Cincinnati, Ohio, June 21. [SRDB Ref ID: 17714]

Hornung, R. W., S. M. Pinney, J. Lodwick, G. Killough, D. E. Brewer, and J. Nasuta, 2008, “Estimation 

of Radon Exposure to Workers at the Fernald Feed Materials Production Center 1952-1988,” 

Journal of Exposure Science and Environmental Epidemiology, 1-12. [SRDB Ref ID: 41619]
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 57 of 76

ICRP (International Commission on Radiological Protection), 1994, Human Respiratory Tract Model 

for Radiological Protection, Publication 66, Pergamon Press, Oxford, England.

Jasailtis, D., and A. Girgzdys, 2007, “Natural Radionuclide Distribution and Radon Exhalation Rate 

from the Soil in Vilnius City,” Journal of Environmental Engineering and Landscape 

Management, volume 15, number 1, pp. 31–37. [SRDB Ref ID: 92984]

Karl, C. L., 1952, “Transfer of Drummed K-65 to the Fernald Area,” memorandum to F. M. Belmore, 

U.S. Atomic Energy Commission, Fernald Area Office, Fernald, Ohio, August 28. [SRDB Ref 

ID: 32897]

LOOW (Lake Ontario Ordnance Works), 1951, “Trip Report and Sample Results,” Memo, W. B. Harris

to F. M. Belmore, “Radiation Exposure Due to Modified Operating Schedule-LOOW”, dated 

12/12/1950, page 33 of 66. [SRDB Ref ID: 8921] 

Mauro, J., 2010, Summary of the Status of DCAS/SC&A Action Items, Fernald WG Section 2.0(3), 

Sanford Cohen and Associates, Vienna, Virginia, September 28. [SRDB Ref ID: 97698] 

Mead, J., F. Savage, and R. Fugate, ca. 1985, History of the Operation of the Feed Materials 

Production Center by NLO, Inc., National Lead Company of Ohio, Cincinnati, Ohio. [SRDB 

Ref ID: 26097]

Methner, M. M., H. A. Feng, and D. F. Utterback, 2001, “Use of Historical Uranium Air Sampling Data 

to Estimate Worker Exposure Potential to Airborne Radioactive Particulate in a Uranium 

Processing Facility,” Applied Occupational and Environmental Hygiene, volume 16, number 

12, pp. 1150–1157. [SRDB Ref ID: 28069]

Morris, R., 2008, FMPC Thorium Timeline, Oak Ridge Associated Universities Team, Oak Ridge, 

Tennessee, February 29. [SRDB Ref ID: 42940]

NIOSH (National Institute for Occupational Safety and Health), 1993, NIOSH Research Issues 

Workshop: Epidemiologic Use of Nondetectable Values in Radiation Exposure 

Measurements, Cincinnati, Ohio, September 9 and 10, 1993, Health-Related Energy Research 

Branch, Cincinnati, Ohio. [SRDB Ref ID: 13723]

NIOSH (National Institute for Occupational Safety and Health), 2004, Technical Information Bulletin: 

Estimation of Ingestion Intakes, OCAS-TIB-009, Rev. 0, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, April 13.

NIOSH (National Institute for Occupational Safety and Health), 2006, SEC Petition Evaluation Report, 

Petition SEC-00046, Final, Office of Compensation Analysis and Support, Cincinnati, Ohio, 

October 25. [SRDB Ref ID: 92983] 

NIOSH (National Institute for Occupational Safety and Health), 2008, Response to SC&A White 

Paper, An Alternative Assessment of Radon Releases from K-65 Silos, Division of 

Compensation Analysis and Support, Cincinnati, Ohio, November. [SRDB Ref ID: 93391] 

NLO (National Lead Company of Ohio), 1957, Exposure Study of Plant 2 Personnel to Airborne 

Radioactive Dust, Health and Safety Division, Survey Section, Cincinnati, Ohio, December 31. 

[SRDB Ref ID: 41028]
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 58 of 76

NLO (National Lead Company of Ohio), 1977, Study of Radioactive Waste Storage Areas at ERDANiagara Falls Site, NLCO-1145, Feed Materials Production Center, Fernald, Ohio, March 25. 

[SRDB Ref ID: 3580]

NLO (National Lead Company of Ohio), 1979, Informational Report on the U.S. Department of 

Energy-Niagara Falls Site, NLCO-003EV, Cincinnati, Ohio, December 1979. [SRDB Ref ID: 

4243]

NRC (National Research Council), 1994, Dose Reconstruction for Fernald Nuclear Facility: A Review 

of Task 4, National Academy Press, Washington, D.C. [SRDB Ref ID: 74362]

ORAUT (Oak Ridge Associated Universities Team), 2004, Technical Basis Document for the Fernald 

Environmental Project (FMPC) – Occupational Internal Dose, ORAUT-TKBS-0017-5, Rev. 00, 

Oak Ridge, Tennessee, May 28. 

ORAUT (Oak Ridge Associated Universities Team), 2005, Estimation of Radium-226 Activity in the 

Body from Breath Radon-222 Measurements, ORAUT-OTIB-0025, Rev. 00, Oak Ridge, 

Tennessee, April 5.

ORAUT (Oak Ridge Associated Universities Team), 2006a, Estimating the Maximum Plausible Dose 

to Workers at Atomic Weapons Employer Facilities, ORAUT-OTIB-0004, Rev. 03 PC-2, Oak 

Ridge, Tennessee, December 6.

ORAUT (Oak Ridge Associated Universities Team), 2006b, Fernald Environmental Management 

Project – Occupational Environmental Dose, ORAUT-TKBS-0017-4, Rev. 01, Oak Ridge, 

Tennessee, February 7. 

ORAUT (Oak Ridge Associated Universities Team), 2007a, Internal Dose Reconstruction, ORAUTOTIB-0060, Rev. 00, Oak Ridge, Tennessee, February 6. 

ORAUT (Oak Ridge Associated Universities Team), 2007b, “FMPC Operations,” documented 

communication with W. Adams et al. (former FMPC employees), Oak Ridge, Tennessee, 

November 9. [SRDB Ref ID: 38124]

ORAUT (Oak Ridge Associated Universities Team), 2010, Hanford Site – Occupational Internal Dose, 

ORAUT-TKBS-0006-5, Rev. 04, Oak Ridge, Tennessee, October 20. 

Pinney, S., R. Hornung, J. Lodwick, D. Brewer, and J. Nasuta, 2004, Condensed Final Report for 

Radon and Cigarette Smoking Exposure Assessment of Fernald Workers, University of 

Cincinnati, RO1 CCR515748-02-01, Department of Environmental Health, College of 

Medicine, and Institute of Health Policies and Health Services, Cincinnati, Ohio, October 22. 

[SRDB Ref ID: 26771]

RAC (Radiological Assessments Corporation), 1995, The Fernald Dosimetry Reconstruction Project, 

Tasks 2 and 3 Radionuclide Source Terms and Uncertainties, CDC-5, Neeses, South 

Carolina, June. [SRDB Ref ID: 3767]

Robinson, 1973, African Metals Corporation Residues, March 6. [SRDB Ref ID: 14800]

Rogers, V. C., and K. K. Nielson, 1991, “Multiphase Radon Generation and Transport in Porous 

Materials, Health Physics, volume 60, pp. 807–815.
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 59 of 76

Ross, K. N., E. D. Leininger, and M. H. Lawrence, 1968, “Exposure Study of Plant 2&3 Personnel to 

Airborne Radioactive Dust,” National Lead of Ohio, Cincinnati, Ohio. [SRDB Ref ID: 41063]

Ross, K. N., E. D. Leininger, and C. W. Zimber, 1969 “Exposure Study of Plant 2&3 Personnel to 

Airborne Radioactive Dust,” March 31, National Lead of Ohio, Cincinnati, Ohio. [SRDB Ref ID: 

41084]

SC&A (Sanford Cohen & Associates), 2008, An Alternative Assessment of Radon Releases from K65 Silos, Vienna, Virginia, November. [SRDB Ref ID: 93391] 

SC&A (Sanford Cohen & Associates), 2010, A Second White Paper Addressing Enhanced Radon 

Releases from the K-65 Silos at the Fernald Site, Vienna, Virginia, April. [SRDB Ref ID: 

93390]

Schumann, C. E., 1952, “K-65 Shipping Operation at L.O.S.A.,” memorandum to R. C. Heatherton, 

National Lead Company of Ohio, Cincinnati, Ohio, November 18. [SRDB Ref ID: 17424]

Scott, L. M., and C. M. West, 1964, Detection and Evaluation of Uranium Exposures, International 

Atomic Energy Agency.

Shaw, W. E, 1968, “Analytical Data of Silo Material,” memorandum to J. E. Carvitti and J. R. Lynch, 

National Lead Company of Ohio, Cincinnati, Ohio, May 24. [SRDB Ref ID: 17612]

SNL (Sandia National Laboratories), 1997, Code Manual for CONTAIN 2.0: A Computer Code for 

Nuclear Reactor Containment Analysis, NUREG/CR-6533, SAND97-1735, U.S. Nuclear 

Regulatory Commission, Washington, D.C. [SRDB Ref ID: 96994] 

Starkey, R. H., 1960–1962, collection of High Air Dust Exposure Operations, National Lead Company 

of Ohio, Feed Materials Processing Center, Fernald, Ohio. [SRDB Ref ID: 2889]

Steckel, L. M. and C. M. West, 1966, Characterization of Y-12 Uranium Process Materials Correlated 

with In Vivo Experience, Y-1544-A. Oak Ridge, Tennessee, July 28. SRDB Ref ID: 11609]

Stefanec, A. J., 1955a, Exposure Study of Plant 2 Personnel to Airborne Radioactive Dust, National 

Lead Company of Ohio, Cincinnati, Ohio, September 22. [SRDB Ref ID: 41017]

Stefanec, A. J., 1955b, Exposure Study of Plant 9 Personnel to Airborne Radioactive Dust, May 17 to 

October 31, 1955, National Lead Company of Ohio, Cincinnati, Ohio, December 30. [SRDB 

Ref ID: 41012]

Stefanec, A. J., and C. R. Schumann, 1956a, Exposure Study of Plant 3 Personnel to Airborne 

Radioactive Dust, National Lead Company of Ohio, Cincinnati, Ohio, October 16. [SRDB Ref 

ID: 41019]

Stefanec, A. J., and C. R. Schumann, 1956b, Exposure Study of Plant 2 Personnel to Airborne 

Radioactive Dust, National Lead Company of Ohio, Health and Safety Division, Cincinnati, 

Ohio, November 2. [SRDB Ref ID: 41020]

Strattman, W. J. 1952, “Review of Production Orders No.41 and 42 (K-65 Handling),” memorandum to 

D. J. Blythe, National Lead Company of Ohio, Cincinnati, Ohio, February 5. [SRDB Ref ID: 

17615]
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 60 of 76

T. Tomes, 2008, Technical Basis for Internal Dosimetry at the FEMP, SD-2008, Rev. 1, Flour Fernald, 

Fernald Closure Project, Fernald, Ohio. [SRDB Ref ID: 3292]

TAC (The Aerospace Corporation), 1982, Background and Resurvey Recommendations for the 

Atomic Energy Commission Portion of the Lake Ontario Ordnance Works, ATR-82(7963-04)-1, 

Washington, D.C., November. [SRDB Ref ID: 12420]

Vogel, R. A., 1989, “The Origin of K-65 Material,” memorandum to P. J. Levine, Westinghouse 

Materials Company, Feed Materials Production Center, Fernald, Ohio, September 18. [SRDB 

Ref ID: 17611]

Walden, C. H., 1952, “K-65 Receipts,” NLO-2458, memorandum to G. W. Wunder, National Lead 

Company of Ohio, Cincinnati, Ohio, April 30. [SRDB Ref ID: 17043, p. 3]

West, C. M., L. M. Scott, and N. B. Schultz, 1979, “Sixteen Years of Uranium Personnel Monitoring 

Experience – In Retrospect,” Health Physics, volume 36. [SRDB Ref ID: 4022, p. 3-7] 

Wing, J. F., 1960, Standard Operating Procedure for Preparing and Issuing Periodic Reports of 

Complete In-Plant Air Dust Surveys, National Lead Company of Ohio, Cincinnati, Ohio, 

December 22. [SRDB Ref ID: 33976, p. 114]. 

Wing, J. F., and R. N. Halcomb, 1958a, Exposure Study of Plant 3 Personnel to Airborne Radioactive 

Dust, National Lead Company of Ohio, Health and Safety Division, Cincinnati, Ohio,

December 4. [SRDB Ref ID: 41050]

Wing, J. F., and R. N. Halcomb, 1958b, Exposure Study of Plant 2 Personnel to Airborne Radioactive 

Dust, National Lead Company of Ohio, Health and Safety Division, Cincinnati, Ohio, 

December 31. [SRDB Ref ID: 41048]

Wing, J. F., and R. N. Halcomb, 1959a, Exposure Study of Plant 2 Personnel to Airborne Radioactive 

Dust, National Lead Company of Ohio, Health and Safety Division, Cincinnati, Ohio, 

December 31. [SRDB Ref ID: 40989]

Wing, J. F., and R. N. Halcomb, 1959b, Exposure Study of Plant 3 Personnel to Airborne Radioactive 

Dust, National Lead Company of Ohio, Health and Safety Division, Cincinnati, Ohio, 

December 31. [SRDB Ref ID: 40990]

Wing, J. F., and E. D. Leininger, 1965, Exposure Study of Plant 2 & 3 Personnel to Airborne 

Radioactive Dust, National Lead Company of Ohio, Health and Safety Division, Cincinnati, 

Ohio. [SRDB Ref ID: 33701}

Ross, K. N., E. D. Leininger, and M. H. Lawrence, 1968, Exposure Study of Plant 2 & 3 Personnel to 

Airborne Radioactive Dust, National Lead of Ohio, Health and Safety Division, Cincinnati, 

Ohio. [SRDB Ref ID: 41063]

Wing, J. F., E. D. Leininger, and R. L. Ruhe, 1966, Exposure Study of Plant 2 & 3 Personnel to 

Airborne Radioactive Dust, National Lead Company of Ohio, Health and Safety Division, 

Cincinnati, Ohio. [SRDB Ref ID: 33702]

Ross, K. N., E. D. Leininger, and C. W. Zimber, 1969, Exposure Study of Plant 2&3 Personnel to 

Airborne Radioactive Dust, National Lead of Ohio, Health and Safety Division, Cincinnati, 

Ohio, March 31. [SRDB Ref ID: 41084]
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 61 of 76

Wing, J. F., K. N. Ross, and G. E. Cline, 1961, Exposure Study of Plant 3 Personnel to Airborne 

Radioactive Dust, National Lead Company of Ohio, Health and Safety Division, Cincinnati, 

Ohio, December 11. [SRDB Ref ID: 41111]

Wing, J. F., K. N. Ross, and E. D. Leininger, 1962, Exposure Study of Plant 2 and 3 Personnel to 

Airborne Radioactive Dust, National Lead Company of Ohio, Health and Safety Division, 

Cincinnati, Ohio, November 27. [SRDB Ref ID: 41126]

Wing, J. F., K. N. Ross, R. L. Ruhe, and R. T. Armbruster, 1961, Exposure Study of Plant 2 Personnel 

to Airborne Radioactive Dust, National Lead Company of Ohio, Health and Safety Division, 

Cincinnati, Ohio, December 11. [SRDB Ref ID: 41110]

Wing, J. F., and R. L. Ruhe, 1960a, Exposure Study of Plant 3 Personnel to Airborne Radioactive 

Dust, National Lead Company of Ohio, Health and Safety Division, Cincinnati, Ohio, 

November 22. [SRDB Ref ID: 41069]

Wing, J. F., and R. L. Ruhe, 1960b, Exposure Study of Plant 2 Personnel to Airborne Radioactive 

Dust, National Lead Company of Ohio, Health and Safety Division, Cincinnati, Ohio, 

December 22. [SRDB Ref ID: 41011]

WMCO (Westinghouse Materials Company of Ohio), 1986,- Memo L.C. Bogar to J. A. Reafsnyder, 

Over-all Accountability Analyses Report, Plant Startup through September 30, 1986, dated 

December 12, 1986. [SRDB Ref ID: 34932]

Wunder, G. W., 1953a, “Resume Activities Month Ending January 20, 1953,” letter to C. L. Carl (U.S. 

Atomic Energy Commission), National Lead Company of Ohio, Cincinnati, Ohio, January 21. 

[SRDB Ref ID: 29525]

Wunder, G. W., 1953b, “Resume of Activities for the Month Ending June 20, 1953,” letter to C. L. Carl 

(U.S. Atomic Energy Commission), National Lead Company of Ohio, Cincinnati, Ohio, June 

19. [SRDB Ref ID: 29522]

Yu, C., C. Loureiro, J.-J. Cheng, L. G. Jones, Y. Y. Wang, Y. P. Chia, and E. Faillace, 1993, Data 

Collection Handbook To Support Modeling Impacts of Radioactive Material in Soil, Argonne 

National Laboratory, Argonne, Illinois, April. [SRDB Ref ID: 93081] 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 62 of 76

Modeling the venting of radon gas from the K-65 Silos was performed by using carbon dioxide as a 

surrogate gas, because it is heavier than air and is in the CONTAIN computer program material 

library, while radon is not. A problem was set up with all cells containing the same atmosphere. The 

pressure reduction for lateral flow was calculated for a wind speed of 6.5 mph. The environment was 

assigned the corresponding reduced initial pressure of 99,420.30 Pa (see Attachment E). The wind 

had little influence on the problem. The internal cells had an initial pressure of 99,422.05 Pa. The 

pressure difference between the silo cells and the environment was eliminated rapidly as little air 

movement was required to achieve equilibrium. All of the internal cells quickly reached the outside 

environmental pressure. Three engineered vents were used to connect the unloading manhole, the 

gooseneck, and the instrument manhole to the environment. The turbulence coefficient used by the 

code was not optimized for the problem. The flow area connecting all internal cells was equal to the 

cross-section of the silo. The flow velocity for the engineered vents was too small for turbulence to be 

important. The principal driver for the problem was heat input to the headspace. 

A heat source of 2,980.77 W was applied to cell 9 for 3,600 seconds (1 hour). The temperature in the 

headspace rose linearly with time from 293 K to 305.42 K. After the heat was turned off at 3,600 

seconds the higher temperature persisted, but the flow rates to the environment fluctuated with 

outflow from the gooseneck, which was partially offset by inflow at the manholes. The calculation was 

run to a problem time of 4 hours. 

For the first problem considered, all the silo cells had an initial atmosphere with 5% CO2 by molar 

fraction. Initially, the flow from the unloading manhole was the largest because of the short path to 

the exterior and the instrument manhole had less area for the flow. The gooseneck path was much 

longer (1.524 m) and had noticeable flow inertia. 

After slightly more than 4 minutes, the inertia was overcome and the gooseneck flow became the 

dominant contributor. Figure A-1 shows the vent flow velocities versus time. After a short time the 

velocities for all three vents are essentially the same. Table A-1 shows the vent velocities versus 

time. 

0.5

0.4

0.3

0.2

0.1

0.0

Vent Flow Velocity (m/sec)

0 200 400 600 800 1000 1200 1400

Time (sec)

Unloading Manhole

Gooseneck

Instrument Manhole

 Problem M2-D

Vent Flow Velocities

Figure A-1. Vent flow velocities versus time. 

ATTACHMENT A

CONTAIN CODE CALCULATIONS AND RESULTS

Page 1 of 7 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 63 of 76

Table A-1. Vent flow velocities versus time. V1 = 

unloading manhole, V2 = gooseneck, V3 = instrument 

manhole. 

Time (s) V1 (m/s) V2 (m/s) V3 (m/s)

10 0.241058 0.096293 0.241061

70 0.175535 0.0904041 0.175538

130 0.406227 0.154696 0.406227

190 0.353121 0.194984 0.353121

250 0.320091 0.21999 0.320091

310 0.299227 0.235722 0.299227

370 0.286029 0.24571 0.286029

430 0.277905 0.252105 0.277905

490 0.272303 0.256172 0.272303

550 0.268871 0.258792 0.268871

610 0.266613 0.26047 0.266613

670 0.265165 0.261547 0.265165

730 0.264231 0.262239 0.264231

790 0.263559 0.262676 0.263559

850 0.263435 0.262977 0.263435

910 0.263014 0.263145 0.263014

970 0.262914 0.263261 0.262914

1,030 0.262823 0.263333 0.262823

1,090 0.262539 0.26336 0.262539

1,150 0.26258 0.263391 0.26258

1,210 0.262457 0.263398 0.262457

1,270 0.26248 0.263407 0.26248

1,330 0.262598 0.263417 0.262598

1,390 0.262578 0.263414 0.262578

1,450 0.26257 0.26341 0.26257

1,510 0.262575 0.263406 0.262575

1,570 0.262631 0.263404 0.262631

Figure A-2 shows the flow rates for the three vents. The total flow to the environment promptly 

reached a value of 17.88 cfm and remained steady thereafter. Table A-2 shows the volumetric flows 

from the engineered vents versus time. 

The CO2 mass in cell 10 (environment) was recorded versus time. Figure A-3 shows and Table A-3 

lists the mass of CO2 vented to the environment for this problem. 

Flow rates were influenced somewhat by the presence of the CO2 compared with plain air 

atmospheres. 

ATTACHMENT A

CONTAIN CODE CALCULATIONS AND RESULTS

Page 2 of 7 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 64 of 76

20

18

16

14

12

10

8

6

4

2

0

Flow to Environment (CFM)

0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 1400 1500

Time (Sec)

Unload Manhole Flow

Gooseneck Flow

Instrument Manhole Flow

Total Flow

 Problem M2-D

All silo cell atmospheres contain

 5% CO2 by molar fraction

Figure A-2. Flows to the environment from the engineered vents. 

Table A-2. Flow from the engineered vents versus time for Problem 

M2-D. Flow1 = unloading manhole, Flow2 = gooseneck, Flow3 = 

instrument manhole. 

Time (s) Flow1 (cfm) Flow2 (cfm) Flow3 (cfm) Total (cfm)

10 6.2764 3.7218 0.79263 10.79083

70 4.5704 3.4942 0.57718 8.64178

130 10.5769 5.9791 1.3357 17.8917

190 9.1941 7.5362 1.1611 17.8914

250 8.3341 8.5027 1.0525 17.8893

310 7.7909 9.1108 0.98388 17.88558

370 7.4473 9.4968 0.94049 17.88459

430 7.2358 9.744 0.91377 17.89357

490 7.0899 9.9012 0.89535 17.88645

550 7.0005 10.0024 0.88407 17.88697

610 6.9417 10.0673 0.87664 17.88564

670 6.904 10.1089 0.87188 17.88478

730 6.8797 10.1357 0.86881 17.88421

790 6.8622 10.1526 0.8667 17.8815

850 6.859 10.1642 0.86619 17.88939

910 6.848 10.1707 0.86481 17.88351

970 6.8454 10.1752 0.86448 17.88508

1,030 6.8431 10.1779 0.86418 17.88515

1,090 6.8357 10.179 0.86325 17.87795

1,150 6.8367 10.1802 0.86338 17.88028

1,210 6.8335 10.1805 0.86298 17.87698

ATTACHMENT A

CONTAIN CODE CALCULATIONS AND RESULTS

Page 3 of 7 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 65 of 76

Time (s) Flow1 (cfm) Flow2 (cfm) Flow3 (cfm) Total (cfm)

1,270 6.8341 10.1808 0.86305 17.87795

1,330 6.8372 10.1812 0.86344 17.88184

1,390 6.8367 10.1811 0.86338 17.88118

1,450 6.8365 10.1809 0.86335 17.88075

1,510 6.8366 10.1808 0.86337 17.88077

1,570 6.8381 10.1807 0.86355 17.88235

3.0

2.5

2.0

1.5

1.0

0.5

0.0

Mass of CO2 Vented to Environment (kg)

14x1030 2 4 6 8 10 12

Time (Sec)

 Problem M2-D

All silo cell atmospheres contain

 5% CO2 by molar fraction

Total initial CO2 mass in silo = 339.9563 kg

Initial CO2 mass in cell 1 = 159.074 kg

Initial CO2 mass in cell 9 = 63.2576 kg

Initial CO2 mass in cells 2-8 = 113.6247 kg

Figure A-3. Mass of CO2 vented to the environment versus time. 

Table A-3. Mass of CO2 vented to the environment versus time for 

Problem M2-D. 

Time (s) CO2 in Cell 10 (kg) Time (s) CO2 in Cell 10 (kg)

10 0.016170 3,900 2.66138

70 0.021368 4,000 2.67535

130 0.0606967 4,300 2.70795

190 0.106139 4,600 2.73212

250 0.151547 4,900 2.75134

310 0.196920 5,000 2.75696

370 0.242260 5,300 2.77205

430 0.287565 5,600 2.78505

490 0.332836 5,900 2.79647

550 0.378073 6,000 2.79999

610 0.423276 6,300 2.80982

670 0.468445 6,600 2.81871

730 0.513580 6,900 2.82683

790 0.558682 7,000 2.82939

850 0.603749 7,600 2.84343

910 0.648783 8,200 2.85565

970 0.693783 8,800 2.86645

1,030 0.738749 9,400 2.87614

ATTACHMENT A

CONTAIN CODE CALCULATIONS AND RESULTS

Page 4 of 7 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 66 of 76

Time (s) CO2 in Cell 10 (kg) Time (s) CO2 in Cell 10 (kg)

1,090 0.783682 10,000 2.88492

1,150 0.828581 10,600 2.89295

1,210 0.873446 11,200 2.90035

1,270 0.918278 11,800 2.90721

1,330 0.963076 12,400 2.91360

1,390 1.00784 13,000 2.91958

1,450 1.05257 13,600 2.92520

1,510 1.09727 14,200 2.93050

1,570 1.14193 14,400 2.93221

1,600 1.16425

1,900 1.38699

2,000 1.46105

2,300 1.68269

2,500 1.82998

2,800 2.05024

3,000 2.19663

3,300 2.41553

3,600 2.59989

The results were divided by the total initial CO2 mass in the problem to give the fraction that was 

vented to the outside. Figure A-4 shows the percent of the CO2 vented to the environment. The 

resulting fraction that was vented at 14,400 seconds (4 hours) was 0.8728%.

1.0

0.8

0.6

0.4

0.2

0.0

% CO2 Vented to the Environment

14x103

0 2 4 6 8 10 12

Time (Sec)

 Problem M2-D

All silo cell atmospheres contain

 5% CO2 by molar fraction

Figure A-4. Percent of the CO2 vented to the environment as a function of time. 

The calculation was repeated with 10%, 15%, and 20% CO2 by molar fraction. Table A-4 shows the 

atmospheric molar fractions for each case. The fraction of the CO2 vented changed only slightly, 

decreasing as the molar fraction increased. Table A-5 shows the final results for each case. The

fraction vented is not sensitive to the molar fraction of CO2 used; it decreases slightly as the molar 

fraction is increased. 

ATTACHMENT A

CONTAIN CODE CALCULATIONS AND RESULTS

Page 5 of 7 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 67 of 76

Table A-4. Atmospheric molar fractions. 

Gas O2 N2 Ar CO2

5% CO2 0.199025 0.741 0.009975 0.05

10% CO2 0.18855 0.702 0.00945 0.1

15% CO2 0.178075 0.663 0.008925 0.15

20% CO2 0.1676 0.624 0.0084 0.2

Table A-5. CO2 vented versus molar fraction CO2. 

Problem M2-D 5% CO2

Time (s) 0.000 14,400

Cell # CO2 Mass (kg) CO2 Mass (kg)

1 159.074 159.072

2 16.2321 16.2319

3 16.2321 16.2319

4 16.2321 16.2319

5 16.2321 16.2319

6 16.2321 16.2319

7 16.2321 16.2319

8 16.2321 16.2318

9 63.2576 60.3289

10 0 2.93221

Total 335.9563 335.95631

% CO2 vented = 0.8728

Problem M2-E 10% CO2

Time (s) 0 14,400

Cell # CO2 Mass (kg) CO2 Mass (kg)

1 318.149 318.145

2 32.4642 32.4638

3 32.4642 32.4638

4 32.4642 32.4638

5 32.4642 32.4638

6 32.4642 32.4638

7 32.4642 32.4638

8 32.4642 32.4636

9 126.515 120.718

10 0 5.80438

Total 671.9134 671.91378

% CO2 vented = 0.8639

ATTACHMENT A

CONTAIN CODE CALCULATIONS AND RESULTS

Page 6 of 7 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 68 of 76

Problem M2-F 15% CO2

Time (s) 0 14,400

Cell # CO2 Mass (kg) CO2 Mass (kg)

1 477.223 477.217

2 48.6963 48.6957

3 48.6963 48.6957

4 48.6963 48.6957

5 48.6963 48.6957

6 48.6963 48.6957

7 48.6963 48.6957

8 48.6963 48.6953

9 189.773 181.165

10 0 8.61876

Total 1007.8701 1007.87026

% CO2 vented = 0.8551

Problem M2-G 20% CO2

Time (s) 0 14,400

Cell # CO2 Mass (kg) CO2 Mass (kg)

1 636.298 636.29

2 64.9284 64.9276

3 64.9284 64.9276

4 64.9284 64.9276

5 64.9284 64.9276

6 64.9284 64.9276

7 64.9284 64.9276

8 64.9284 64.9266

9 253.03 241.668

10 0 11.3774

Total 1343.8268 1343.8276

% CO2 vented = 0.8466

ATTACHMENT A

CONTAIN CODE CALCULATIONS AND RESULTS

Page 7 of 7 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 69 of 76

Gray (1963), pp. 4-103, -108, -114,

CO2 

Density at 300 K, 1 atmosphere 3 3

0

0

3 3

1.977 10 /

0.9088 @ 300 K, 1 atm

1.796697 10 /

g cm

g cm

ρ

ρ

ρ

ρ

−

−

= ×

=

= ×

Specific Heat CP

C R P = 4.513 @ 300K,1 atm Reference: 0.188921 J/gK

CP = 0.852600 J/gK

Air

3 3

0

0

3 3

1.29304 10 /

0.9102 @ 300 K, 1 atm

1.1769 10 /

g cm

g cm

ρ

ρ

ρ

ρ

−

−

= ×

=

= ×

Specific Heat CP

C R P = 3.5059 @ 300K,1 atm Reference: 0.287041 J/gK

CP = 1.006337 J/gK

ATTACHMENT B

GAS DATA APPLIED IN CONTAIN MODEL
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 70 of 76

The CONTAIN model includes wind pressure as another driving force for transporting the radon (CO2) 

to the environment. 

Wind pressure effects are evaluated using the American Society of Heating, Refrigeration, and AirConditioning Engineers methodology and are applied to the model environment cells. The building 

wind pressure is calculated from the dynamic head equation written in the form:

2

0.0129

2

w

w w

v ∆ = P c ρ

where:

 ∆ = Pw Localized air pressure change due to wind (inches water) 

 wc = Wind pressure coefficient (dimensionless)

 wv = Wind speed (mph)

ρ = Air density (lb/ft3

) 

The wind coefficient wc varies with wind direction in relation to building surfaces. With wind impinging 

normally on a wall, cw = +0.7 . Wind parallel to a wall produces a wind coefficient of –0.35 and on the 

downwind side of the building a coefficient of –0.4 is used. With these wind coefficients, the pressure 

difference across the facility was calculated and analysis was performed for each wind direction.

ATTACHMENT C

WIND MODELING IN THE CONTAIN MODEL
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 71 of 76

Unloading Manhole 

Path length = 4” =0.1016 m

( )

( )

2

2

2

2

2

1

12.25 0.3048 / 0.3041512

12 /

12.00 0.3048 / 0.2918635

12 /

in A m ft m

in ft

in A m ft m

in ft

π

π

  = =    

  = =    

Gap Area = 2 2 A A 2 1 1.22877 10 m− −= × AL m / 0.120942 =

Gooseneck

Path length = 5 ft = 1.524 m

Diameter = 6 in., Radius = 3 in.

( )

2 2 2 A ft m ft 0.25 0.3048 / 1.8241 10 m− = × =× AL m / 0.011969 =

Instrument manhole

Path length = 4 in. =0.1016 m

( )

( )

2

2 2

2

2

2 2

1

3.125 0.3048 / 1.979326 10

12 /

3.00 0.3048 / 1.8241468 10

12 /

in A m ft m

in ft

in A m ft m

in ft

π

π

−

−

  = = ×    

  = = ×    

Gap Area = 3 2 A A 2 1 1.551792 10 m− −= × AL m / 0.0152735 =

ATTACHMENT D

VENT GEOMETRY CALCULATIONS
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 72 of 76

&& Model #M2-D for K-65 Silo Simulation 

&&

&&

control

ncells = 10 

ntitl = 11

ntzone = 10

numtbg = 2

maxtbg = 50

nengv = 3

nmtrac = 1

eoi

&&

material

compound n2 o2 ar conc h2ov ss co2

trackmat co2

eoi

&& -----------------Title-----------------

title 

 K-65 Silo Airflow Simulation.

The domed headspace is simulated by a cylinder with equal volume.

Cell 1 contains the debris - assumed to have 0.7porosity. The 

volume of the cell counts only the portion available to gas. 

The connections of cell 9 (headspace) to the outside environment 

are by three engineered vents: one for the Unloading Manhole, one

for the Gooseneck, and one for the Instrument Manhole. Cells 2 

through 8 are for observing gradients in the gas. All internal 

cells contain air with 0.05 molar fraction of CO2. CO2 is tracked

in the problem. The environmental cell contains only air. 

&&

&&

&& -----------------Edit Times------------

times 30000.0 0.0

0.1 5. 10.0

1.0 20. 1600.0

1.0 100. 2000.0

1.0 100. 2500.0

1.0 100. 3000.0

1.0 100. 4000.0

1.0 100. 5000.0

1.0 100. 6000.0

2.0 100. 7000.0

4.0 200. 14400.0

&& ----------------Specify Thermal Reactor---

Thermal

&& -----------------Output frequency---------

ATTACHMENT E

PROBLEM INPUT DECK

Page 1 of 5 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 73 of 76

shortedt = 3

&& ---Long edit output variables desired ----

longedt = 3

&& ----------------Flow Data----------------- 

flows

implicit

perror = 0.1

&& ----------------Flow paths ---------------

&& Turbulent flow coefficients (CFC) for closed hatches

&& were calculated using the methodology found in

&& CALC-906-NA-000013 by Robert Colwell. For open paths

&& a value of CFC = 0.01 was assigned since turbulence for 

&& flows through large openings should be minor. 

&& -------Internal Layers of the Silo-----------

area(1,2) = 466.98 && (silo cross section) bottom layer to second layer

avl(1,2) = 1556.6 && Assumes 0.3m path length

cfc(1,2) = 0.01 && Arbitrary choice for an open path

topen(1,2) = 0.0

area(2,3) = 466.98 && (silo cross section) second layer to third layer

avl(2,3) = 1556.6 && Assumes 0.3m path length

cfc(2,3) = 0.01 && Arbitrary choice for an open path

topen(2,3) = 0.0

area(3,4) = 466.98 && (silo cross section) third layer to fourth layer

avl(3,4) = 1556.6 && Assumes 0.3m path length

cfc(3,4) = 0.01 && Arbitrary choice for an open path

topen(3,4) = 0.0

area(4,5) = 466.98 && (silo cross section) fourth layer to fifth layer

avl(4,5) = 1556.6 && Assumes 0.3m path length 

cfc(4,5) = 0.01 && Arbitrary choice for an open path

topen(4,5) = 0.0

area(5,6) = 466.98 && (silo cross section) fifth layer to sixth layer

avl(5,6) = 1556.6 && Assumes 0.3m path length

cfc(5,6) = 0.01 && Arbitrary choice for an open path

topen(5,6) = 0.0

area(6,7) = 466.98 && (silo cross section) sixth layer to seventh layer

avl(6,7) = 1556.6 && Assumes 0.3m path length

cfc(6,7) = 0.01 && Arbitrary choice for an open path

topen(6,7) = 0.0

area(7,8) = 466.98 && (silo cross section) seventh layer to eighth layer 

ATTACHMENT A

Page 1 of 5 

ATTACHMENT E

PROBLEM INPUT DECK

Page 2 of 5 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 74 of 76

avl(7,8) = 1556.6 && Assumes 0.3m path length 

cfc(7,8) = 0.01 && Arbitrary choice for an open path

topen(7,8) = 0.0

area(8,9) = 466.98 && (silo cross section) eighth layer to ninth layer

avl(8,9) = 1556.6 && Assumes 0.3m path length

cfc(8,9) = 0.01 && Arbitrary choice for an open path 

topen(8,9) = 0.0

engvent

from = 9 to = 10

varea = 1.2288e-02 && (Unloading Manhole) ninth layer to exterior

vavl = 0.12094 && Assumes 0.1016m path length

vcfc = 0.01 && Arbitrary choice for an open path

vtopen = 0.0

eoi

from = 9 to = 10

varea = 1.8241e-02 && (Gooseneck) ninth layer to exterior

vavl = 0.011969 && Assumes 1.524m path length

vcfc = 0.01 && Arbitrary choice for an open path

vtopen = 0.0

eoi

from = 9 to = 10

varea = 1.5518e-03 && (Instrument Manhole) ninth layer to exterior

vavl = 0.0152735 && Assumes 0.1016m path length

vcfc = 0.01 && Arbitrary choice for an open path

vtopen = 0.0

eoi

&& -------Print optional outputs----------------

prflow

&& ----------Begin cell inputs------------------

cell = 1 && Bottom layer containing radioactive material

control 

eoi

geometry 1771.29 3.7931 && (gas volume only - ignores solids)

&& Assumes cylinder portion of silo is 2/3 full of material

atmos = 4 99422.05 293.0

o2 = 0.199025 n2 = 0.741 ar = 0.009975 co2 = 0.05

eoi

&& ------------End of cell 1 inputs

cell = 2 && Layer 2 (first layer above radioactive material)

control

eoi

ATTACHMENT E

PROBLEM INPUT DECK

Page 3 of 5 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 75 of 76

geometry 180.744 0.38705 

atmos = 4 99422.05 293.0

o2 = 0.199025 n2 = 0.741 ar = 0.009975 co2 = 0.05

eoi

&& ------------End of cell 2 inputs----------

cell = 3 && layer 3 (second layer above radioactive material - all gas)

control

eoi

geometry 180.744 0.38705 

atmos = 4 99422.05 293.0

o2 = 0.199025 n2 = 0.741 ar = 0.009975 co2 = 0.05

eoi

&& ------------End of cell 3 inputs----------

cell = 4 && layer 4 (third layer above radioactive material - all gas)

control

eoi

geometry 180.744 0.380705 

atmos = 4 99422.05 293.0

o2 = 0.199025 n2 = 0.741 ar = 0.009975 co2 = 0.05

eoi

&& ------------End of cell 4 inputs----------

cell = 5 && layer 5 (fourth layer above radioactive material - all gas)

control

eoi

geometry 180.744 0.38705 

atmos = 4 99422.05 293.0

o2 = 0.199025 n2 = 0.741 ar = 0.009975 co2 = 0.05

eoi

&& ------------End of cell 5 inputs----------

cell = 6 && layer 6 (fifth layer above radioactive material - all gas)

control

eoi

geometry 180.744 0.38705 

atmos = 4 99422.05 293.0

o2 = 0.199025 n2 = 0.741 ar = 0.009975 co2 = 0.05

eoi

&& ------------End of cell 6 inputs----------

cell = 7 && layer 7 (sixth layer above radioactive material - all gas)

control

eoi

geometry 180.744 0.38705 

atmos = 4 99422.05 293.0

o2 = 0.199025 n2 = 0.741 ar = 0.009975 co2 = 0.05

eoi

&& ------------End of cell 7 inputs----------

ATTACHMENT E

PROBLEM INPUT DECK

Page 4 of 5 
Document No. ORAUT-RPRT-0052 Revision No. 00 Effective Date: 07/12/2011 Page 76 of 76

cell = 8 && layer 8 (seventh layer above radioactive material - all gas)

control

eoi

geometry 180.744 0.38705 

atmos = 4 99422.05 293.0

o2 = 0.199025 n2 = 0.741 ar = 0.009975 co2 = 0.05

eoi

&& ------------End of cell 8 inputs----------

cell = 9 && layer 9 (Head Space - eighth layer above radioactive material)

control

nsoatm = 1

nspatm = 50

numtbc = 1

maxtbc = 50

eoi

geometry 704.372 1.5084 && (equivalent cylinder)

atmos = 4 99422.05 293.0

o2 = 0.199025 n2 = 0.741 ar = 0.009975 co2 = 0.05

eoi

&& ----------2.98077E-03 Megawatt Fire Simulation--------

source = 1

h2ov = 5

iflag = 2

t = 0.0 100. 3500. 3600. 4000.

mass = 0.0 1.0e-10 1.0e-10 0.0 0.0

enth = 0.0 2.98077e+13 2.98077e+13 0.0 0.0

eoi

&& ------------End of cell 9 inputs----------

cell = 10 && Outside Atmosphere 

control

eoi

geometry 1.0e10 20.

atmos = 3 99420.30 293.0 && 1.751 Pa below ambient (lateral wind)

o2 = 0.2095 n2 = 0.78 ar = 0.0105

eoi

&& ------------End of cell 10 inputs----------

eof

ATTACHMENT E

PROBLEM INPUT DECK

Page 5 of 5